ML18022A443

Nine Mile Point, Unit 1, UFSAR, Rev 25 - Redacted

ML18022A443
Person / Time
Site:	Nine Mile Point
Issue date:	01/22/2018
From:	Niagara Mohawk Power Corp
To:	Marshall M L Plant Licensing Branch 1
Marhsall M L, NRR/DORL/LPL, 415-2871
Shared Package
ML17286A075	List: ML18022A443
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Download: ML18022A443 (1548)

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{{#Wiki_filter:NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title TOC i Rev. 25, October 2017 TABLE OF CONTENTS LIST OF TABLES LIST OF FIGURES SECTION I INTRODUCTION AND SUMMARY A. PRINCIPAL DESIGN CRITERIA 1.0 General 2.0 Buildings and Structures 3.0 Reactor 4.0 Reactor Vessel 5.0 Containment 6.0 Control and Instrumentation 7.0 Electrical Power 8.0 Radioactive Waste Disposal 9.0 Shielding and Access Control 10.0 Fuel Handling and Storage B. CHARACTERISTICS 1.0 Site 2.0 Reactor 3.0 Core 4.0 Fuel Assembly 5.0 Control System 6.0 Core Design and Operating Conditions 7.0 Design Power Peaking Factor 8.0 Nuclear Design Data 9.0 Reactor Vessel 10.0 Coolant Recirculation Loops 11.0 Primary Containment 12.0 Secondary Containment 13.0 Structural Design 14.0 Station Electrical System 15.0 Reactor Instrumentation System 16.0 Reactor Protection System C. IDENTIFICATION OF CONTRACTORS D. GENERAL CONCLUSIONS NMP Unit 1 UFSAR Section Title TOC ii Rev. 25, October 2017 E. REFERENCES SECTION II STATION SITE AND ENVIRONMENT A. SITE DESCRIPTION 1.0 General 2.0 Physical Features 3.0 Property Use and Development B. DESCRIPTION OF AREA ADJACENT TO THE SITE 1.0 General 1.1 Population 2.0 Agriculture, Industrial and Recreational Use 2.1 Agricultural Use 2.2 Industrial Use 2.2.1 Toxic Chemicals 2.3 Recreational Use C. METEOROLOGY D. LIMNOLOGY E. EARTH SCIENCES F. ENVIRONMENTAL RADIOLOGY G. REFERENCES SECTION III BUILDINGS AND STRUCTURES A. TURBINE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating and Ventilation System 2.3 Smoke and Heat Removal NMP Unit 1 UFSAR Section Title TOC iii Rev. 25, October 2017 2.4 Shielding and Access Control 2.5 Additional Building Cooling 3.0 Safety Analysis B. CONTROL ROOM 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating, Ventilation and Air Conditioning System 2.3 Smoke and Heat Removal 2.4 Shielding and Access Control 3.0 Safety Analysis C. WASTE DISPOSAL BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating and Ventilation System 2.3 Shielding and Access Control 3.0 Safety Analysis D. OFFGAS BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating and Ventilation System 2.3 Shielding and Access Control NMP Unit 1 UFSAR Section Title TOC iv Rev. 25, October 2017 3.0 Safety Analysis E. NONCONTROLLED BUILDINGS 1.0 Administration Building 1.1 Design Bases 1.1.1 Wind and Snow Loadings 1.1.2 Pressure Relief Design 1.1.3 Seismic Design and Internal Loadings 1.1.4 Heating, Cooling and Ventilation 1.1.5 Shielding and Access Control 1.2 Structure Design 1.2.1 General Structural Features 1.2.2 Heating, Ventilation and Air Conditioning 1.2.3 Access Control 1.3 Safety Analysis 2.0 Sewage Treatment Building 2.1 Design Bases 2.1.1 Wind and Snow Loadings 2.1.2 Pressure Relief Design 2.1.3 Seismic Design and Internal Loadings 2.1.4 Electrical Design 2.1.5 Fire and Explosive Gas Detection 2.1.6 Heating and Ventilation 2.1.7 Shielding and Access Control 2.2 Structure Design 2.2.1 General Structural Features 2.2.2 Ventilation System 2.2.3 Access Control 3.0 Energy Information Center 3.1 Design Bases 3.1.1 Wind and Snow Loadings 3.1.2 Pressure Relief Design 3.1.3 Seismic Design and Internal Loadings 3.1.4 Heating and Ventilation 3.1.5 Shielding and Access Control 3.2 Structure Design 3.2.1 General Structural Features 3.2.2 Heating and Ventilation System 3.2.3 Access Control F. SCREENHOUSE, INTAKE AND DISCHARGE TUNNELS NMP Unit 1 UFSAR Section Title TOC v Rev. 25, October 2017 1.0 Screenhouse 1.1 Design Basis 1.1.1 Wind and Snow Loadings 1.1.2 Pressure Relief Design 1.1.3 Seismic Design and Internal Loadings 1.1.4 Heating and Ventilation 1.1.5 Shielding and Access Control 1.2 Structure Design 2.0 Intake and Discharge Tunnels 2.1 Design Bases 2.2 Structure Design 3.0 Safety Analysis G. STACK 1.0 Design Bases 1.1 General 1.2 Wind Loading 1.3 Seismic Design 1.4 Shielding and Access Control 2.0 Structure Design 3.0 Safety Analysis 3.1 Radiology 3.2 Stack Failure Analysis 3.2.1 Reactor Building 3.2.2 Diesel Generator Building 3.2.3 Screen and Pump House H. SECURITY BUILDING WEST AND SECURITY BUILDING ANNEX I. RADWASTE SOLIDIFICATION AND STORAGE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating, Ventilation and Air Conditioning 1.5 Shielding and Access Control 2.0 Structure and Design 2.1 General Structural Features 2.2 Heating, Ventilation and Air NMP Unit 1 UFSAR Section Title TOC vi Rev. 25, October 2017 Conditioning 2.3 Shielding and Access Control 3.0 Use J. REFERENCES SECTION IV REACTOR A. DESIGN BASES 1.0 General 2.0 Performance Objectives 3.0 Design Limits and Targets B. REACTOR DESIGN 1.0 General 2.0 Nuclear Design Technique 2.1 Reference Loading Pattern 2.2 Final Loading Pattern 2.2.1 Acceptable Deviation From Reference Loading Pattern 2.2.2 Reexamination of Licensing Basis 2.3 Refueling Cycle Reactivity Balance 3.0 Thermal and Hydraulic Characteristics 3.1 Thermal and Hydraulic Design 3.1.1 Recirculation Flow Control 3.1.2 Core Thermal Limits 3.1.2.1 Excessive Clad Temperature 3.1.2.2 Cladding Strain 3.1.2.3 Coolant Flow 3.2 Thermal and Hydraulic Analyses 3.2.1 Hydraulic Analysis 3.2.2 Thermal Analysis 3.2.2.1 Fuel Cladding Integrity Safety Limit Analysis 3.2.2.2 MCPR Operating Limit Analysis 3.3 Reactor Transients 4.0 Stability Analysis 4.1 Design Bases 4.2 Stability Analysis Method 5.0 Mechanical Design and Evaluation 5.1 Fuel Mechanical Design 5.1.1 Design Bases 5.1.2 Fuel Rods NMP Unit 1 UFSAR Section Title TOC vii Rev. 25, October 2017 5.1.3 Water Rods 5.1.4 Fuel Assemblies 5.1.5 Mechanical Design Limits and Stress Analysis 5.1.6 Relationship Between Fuel Design Limits and Fuel Damage Limits 5.1.7 Surveillance and Testing 6.0 Control Rod Mechanical Design and Evaluation 6.1 Design 6.1.1 Control Rods and Drives 6.1.2 Standby Liquid Poison System 6.2 Control System Evaluation 6.2.1 Rod Withdrawal Errors Evaluation 6.2.2 Overall Control System Evaluation 6.3 Limiting Conditions for Operation and Surveillance 6.4 Control Rod Lifetime 7.0 Reactor Vessel Internal Structure 7.1 Design Bases 7.1.1 Core Shroud 7.1.2 Core Support 7.1.3 Top Grid 7.1.4 Control Rod Guide Tubes 7.1.5 Feedwater Sparger 7.1.6 Core Spray Spargers 7.1.7 Liquid Poison Sparger 7.1.8 Steam Separator and Dryer 7.1.9 Core Shroud Stabilizers 7.1.10 Core Shroud Vertical Weld Repair 7.2 Design Evaluation 7.3 Surveillance and Testing C. REFERENCES SECTION V REACTOR COOLANT SYSTEM A. DESIGN BASES 1.0 General 2.0 Performance Objectives 3.0 Design Pressure 4.0 Cyclic Loads (Mechanical and Thermal) 5.0 Codes NMP Unit 1 UFSAR Section Title TOC viii Rev. 25, October 2017 B. SYSTEM DESIGN AND OPERATION 1.0 General 1.1 Drawings 1.2 Materials of Construction 1.3 Thermal Stresses 1.4 Primary Coolant Leakage 1.5 Coolant Chemistry 2.0 Reactor Vessel 3.0 Reactor Recirculation Loops 4.0 Reactor Steam and Auxiliary Systems Piping 5.0 Relief Devices C. SYSTEM DESIGN EVALUATION 1.0 General 2.0 Pressure 3.0 Design Heatup and Cooldown Rates 4.0 Materials Radiation Exposure 4.1 Pressure-Temperature Limit Curves 4.2 Temperature Limits for Boltup 4.3 Temperature Limits for In-Service System Pressure Tests 4.4 Operating Limits During Heatup, Cooldown, and Core Operation 4.5 Predicted Shift in RTNDT 4.6 Neutron Fluence Calculations 5.0 Mechanical Considerations 5.1 Jet Reaction Forces 5.2 Seismic Forces 6.0 Safety Limits, Limiting Safety Settings and Minimum Conditions for Operation D. TESTS AND INSPECTIONS 1.0 Prestartup Testing 2.0 Inspection and Testing Following Startup 2.1 Pressure Test 2.2 Pressure Vessel Irradiation E. EMERGENCY COOLING SYSTEM 1.0 Design Bases 2.0 System Design and Operation NMP Unit 1 UFSAR Section Title TOC ix Rev. 25, October 2017 3.0 Design Evaluation 3.1 Redundancy 3.2 Makeup Water 3.3 System Leaks 3.4 Containment Isolation 4.0 Tests and Inspections 4.1 Prestartup Test 4.2 Subsequent Inspections and Tests F. REFERENCES SECTION VI CONTAINMENT SYSTEM A. PRIMARY CONTAINMENT - MARK I CONTAINMENT PROGRAM 1.0 General Structure 2.0 Pressure Suppression Hydrodynamic Loads 2.1 Safety/Relief Valve Discharge 2.2 Loss-of-Coolant Accident 2.3 Summary of Loading Phenomena 3.0 Plant-Unique Modifications B. PRIMARY CONTAINMENT - PRESSURE SUPPRESSION SYSTEM 1.0 Design Bases 1.1 General 1.2 Design Basis Accident (DBA) 1.3 Containment Heat Removal 1.4 Isolation Criteria 1.5 Vacuum Relief Criteria 1.6 Flooding Criteria 1.7 Shielding 2.0 Structure Design 2.1 General 2.2 Penetrations and Access Openings 2.3 Jet and Missile Protection 2.4 Materials 2.5 Shielding 2.6 Vacuum Relief 2.7 Containment Flooding C. SECONDARY CONTAINMENT - REACTOR BUILDING NMP Unit 1 UFSAR Section Title TOC x Rev. 25, October 2017 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design 1.4 Shielding 2.0 Structure Design 2.1 General Structural Features D. CONTAINMENT ISOLATION SYSTEM 1.0 Design Bases 1.1 Containment Spray Appendix J Water Seal Requirements 2.0 System Design 3.0 Tests and Inspections E. CONTAINMENT VENTILATION SYSTEM 1.0 Primary Containment 1.1 Design Bases 1.2 System Design 2.0 Secondary Containment 2.1 Design Bases 2.2 System Design F. TEST AND INSPECTIONS 1.0 Drywell and Suppression Chamber 1.1 Preoperational Testing 1.2 Postoperational Testing 2.0 Containment Penetrations and Isolation Valves 2.1 Penetration and Valve Leakage 2.2 Valve Operability Test 3.0 Containment Ventilation System 4.0 Other Containment Tests 5.0 Reactor Building 5.1 Reactor Building Normal Ventilation System 5.2 Reactor Building Isolation Valves 5.3 Emergency Ventilation System G. REFERENCES SECTION VII ENGINEERED SAFEGUARDS A. CORE SPRAY SYSTEM NMP Unit 1 UFSAR Section Title TOC xi Rev. 25, October 2017 1.0 Design Bases 2.0 System Design 2.1 General 2.2 Operator Assessment 3.0 Design Evaluation 4.0 Tests and Inspections B. CONTAINMENT SPRAY SYSTEM 1.0 Licensing Basis Requirements 1.1 10CFR50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 1.2 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants 2.0 Design Bases 2.1 Design Basis Functional Requirements 2.2 Controlling Parameters 3.0 System Design 3.1 System Function 3.2 System Design Description 3.3 System Design 3.4 Codes and Standards 3.5 System Instrumentation 3.6 System Design Features 4.0 Design Performance Evaluation 4.1 System Performance Analyses 4.2 System Response 4.3 Interdependency With Other Engineered Safeguards Systems 5.0 System Operation 5.1 Limiting Conditions for Operation 6.0 Tests and Inspection C. LIQUID POISON INJECTION SYSTEM 1.0 Design Bases 2.0 System Design 2.1 Operator Assessment 3.0 Design Evaluation 4.0 Tests and Inspections 5.0 Alternate Boron Injection D. CONTROL ROD VELOCITY LIMITER 1.0 Design Bases NMP Unit 1 UFSAR Section Title TOC xii Rev. 25, October 2017 2.0 System Design 3.0 Design Evaluation 3.1 General 3.2 Design Sensitivity 3.3 Normal Operation 4.0 Tests and Inspections E. CONTROL ROD HOUSING SUPPORT 1.0 Design Bases 2.0 System Design 2.1 Loads and Deflections 3.0 Design Evaluation 4.0 Tests and Inspections F. FLOW RESTRICTORS 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections G. COMBUSTIBLE GAS CONTROL SYSTEM 1.0 Design Bases 2.0 Containment Inerting System 2.1 System Design 2.2 Design Evaluation 3.0 Containment Atmospheric Dilution System 3.1 System Design 3.2 Design Evaluation 4.0 Tests and Inspections H. EMERGENCY VENTILATION SYSTEM 1.0 Design Bases 2.0 System Design 2.1 Operator Assessment 3.0 Design Evaluation 4.0 Tests and Inspections I. HIGH-PRESSURE COOLANT INJECTION 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections NMP Unit 1 UFSAR Section Title TOC xiii Rev. 25, October 2017 J. GAS ACCUMULATION 1.0 Nine Mile Point Response to GL 2008-01 K. REFERENCES SECTION VIII INSTRUMENTATION AND CONTROL A. PROTECTIVE SYSTEMS 1.0 Design Bases 1.1 Reactor Protection System 1.2 Anticipated Transients Without Scram Mitigation System 2.0 System Design 2.1 Reactor Protection System 2.2 Anticipated Transients Without Scram Mitigation System 3.0 System Evaluation B. REGULATING SYSTEMS 1.0 Design Bases 2.0 System Design 2.1 Control Rod Adjustment Control 2.2 Recirculation Flow Control 2.3 Pressure and Turbine Control 2.4 Reactor Feedwater Control 3.0 System Evaluation 3.1 Control Rod Adjustment Control 3.2 Recirculation Flow Control 3.3 Pressure and Turbine Control 3.4 Reactor Feedwater Control C. INSTRUMENTATION SYSTEMS 1.0 Nuclear Instrumentation 1.1 Design 1.1.1 Source Range Monitors 1.1.2 Intermediate Range Monitors 1.1.3 Local Power Range Monitors 1.1.4 Average Power Range Monitors 1.1.5 Traversing In-Core Probe System 1.2 Evaluation 1.2.1 Source Range Monitors 1.2.2 Intermediate Range Monitors 1.2.3 Local Power Range Monitors 1.2.4 Average Power Range Monitors NMP Unit 1 UFSAR Section Title TOC xiv Rev. 25, October 2017 2.0 Nonnuclear Process Instrumentation 2.1 Design Bases 2.1.1 Nonnuclear Process Instruments in Protective Systems 2.1.2 Nonnuclear Process Instruments in Regulating Systems 2.1.3 Other Nonnuclear Process Instruments 2.2 Evaluation 2.2.1 Nonnuclear Process Instruments in Protective System 2.2.2 Nonnuclear Process Instruments in Regulating Systems 2.2.3 Other Nonnuclear Process Instruments 3.0 Radioactivity Instrumentation 3.1 Design Bases 3.1.1 Radiation Monitors in Protective Systems 3.1.2 Other Radiation Monitors 3.2 Evaluation 4.0 Other Instrumentation 4.1 Rod Worth Minimizer 4.1.1 Design Bases 4.1.2 Evaluation 4.2 Offgas System Explosive Gas Monitoring 4.2.1 Design Bases 4.2.2 Surveillance Requirements 4.2.2.1 Hydrogen Monitor Operability Demonstration 4.2.2.2 Hydrogen Concentration Requirement 5.0 Regulatory Guide 1.97 (Revision 2) Instrumentation 5.1 Licensing Activities - Background 5.2 Definition of RG 1.97 Variable Types and Instrument Categories 5.3 Determination of RG 1.97 Type A Variables for Unit 1 5.4 Determination of EOP Key Parameters for Unit 1 5.4.1 Determination Basis/Approach 5.4.2 Definition of Primary Safety Functions NMP Unit 1 UFSAR Section Title TOC xv Rev. 25, October 2017 5.4.3 Association of EOPs to Primary Safety Functions 5.4.4 Identification of EOP Key Parameters 5.5 Unit 1 RG 1.97 Variables, Variable Type, and Associated Instrument Category Designations 5.6 Summary of the RG 1.97 Instrument Design and Implementation Criteria that were Established for Unit 1 as Part of the Unit 1 1990 Restart Activities 5.6.1 No Type A Variables 5.6.2 EOP Key Parameters 5.6.3 Single Tap for the Fuel Zone RPV Water Level Instrument 5.6.4 Nonredundant Wide-Range RPV Water Level Indication 5.6.5 Upgrading EOP Key Parameter Category 1 Instrument Loop Components to Safety-Related Classification 5.6.6 Safety-Related Classification of Instrumentation for RG 1.97 Variable Types other than the EOP Key Parameters 5.6.7 Routing and Separation of Channelized Category 1 Instrument Loop Cables 5.6.8 Electrical Isolation of Category 1 Instrument Loops from Associated Components that are not Safety Related 5.6.9 Power Source Information for Category 1 Instruments 5.6.10 Marking of Instruments of Control Room Panels 5.6.11 "Alternate" Instruments for Monitoring EOP Key Parameters 5.6.12 Indication Ranges of Monitoring Instruments D. REFERENCES NMP Unit 1 UFSAR Section Title TOC xvi Rev. 25, October 2017 SECTION IX ELECTRICAL SYSTEMS A. DESIGN BASES B. ELECTRICAL SYSTEM DESIGN 1.0 Network Interconnections 1.1 345-kV System 1.2 115-kV System 2.0 Station Distribution System 2.1 Two 24-V Dc Systems 2.2 Two 120-V, 60-Hz, Single-Phase, Uninterruptible Power Supply Systems 2.3 Two 120-V, 57-60 Hz, One-Phase, Reactor Trip Power Supplies 2.4 One 120/208-V, 60-Hz, Instrument and Control Transformer 2.5 One 120/240-V, 60-Hz, Single-Phase, Computer Power Supply 3.0 Cables and Cable Trays 3.1 Cable Separation 3.2 Cable Penetrations 3.3 Protection in Hazardous Areas 3.4 Types of Cables 3.4.1 Power Cable 3.4.2 Control Cable 3.4.3 Special Cable 3.5 Design and Spacing of Cable Trays 3.5.1 Tray Design Specifications 3.5.2 Tray Spacing 4.0 Emergency Power 4.1 Diesel Generator System 4.2 Station Batteries 4.3 Q-Related Battery System 5.0 Tests and Inspections 5.1 Diesel Generator 5.2 Station Batteries 5.3 Q-Related Battery 6.0 Conformance With 10CFR50.63 - Station Blackout Rule 6.1 Station Blackout Duration 6.2 Station Blackout Coping Capability 6.3 Procedures and Training 6.4 Quality Assurance 6.5 Emergency Diesel Generator NMP Unit 1 UFSAR Section Title TOC xvii Rev. 25, October 2017 Reliability Program 6.6 References SECTION X REACTOR AUXILIARY AND EMERGENCY SYSTEMS A. REACTOR SHUTDOWN COOLING SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections B. REACTOR CLEANUP SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections C. CONTROL ROD DRIVE HYDRAULIC SYSTEM 1.0 Design Bases 2.0 System Design 2.1 Pumps 2.2 Filters 2.3 First Pressure Stage 2.4 Second Pressure Stage 2.5 Third Pressure Stage 2.6 Exhaust Header 2.7 Accumulator 2.8 Scram Pilot Valves 2.9 Scram Valves 2.10 Scram Dump Volume 2.11 Control Rod Drive Cooling System 2.12 Directional Control and Speed Control Valves 2.13 Rod Insertion and Withdrawal 2.14 Scram Actuation 3.0 System Evaluation 3.1 Normal Withdrawal Speed 3.2 Accidental Multiple Operation 3.3 Scram Reliability 3.4 Operational Reliability 3.5 Alternate Rod Injection 4.0 Reactor Vessel Level Instrumentation Reference Leg Backfill NMP Unit 1 UFSAR Section Title TOC xviii Rev. 25, October 2017 5.0 Tests and Inspections D. REACTOR BUILDING CLOSED LOOP COOLING WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections E. TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections F. SERVICE WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections G. MAKEUP WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections H. SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections I. BREATHING, INSTRUMENT AND SERVICE AIR SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections J. FUEL AND REACTOR COMPONENTS HANDLING SYSTEM NMP Unit 1 UFSAR Section Title TOC xix Rev. 25, October 2017 1.0 Design Bases 2.0 System Design 2.1 Description of Facility 2.1.1 Cask Drop Protection System 2.2 Operation of the Facility 2.3 Control of Heavy Loads Program 2.3.1 Introduction/Licensing Background 2.3.2 Safety Basis 2.3.3 Scope of Heavy Load Handling Systems 2.3.4 Control of Heavy Loads Program 2.3.4.1 NMPNS Commitments in Response to NUREG-0612, Phase I Elements 2.3.4.2 Reactor Pressure Vessel Head and Spent Fuel Cask Lifts 2.3.5 Safety Evaluation 3.0 Design Evaluation 4.0 Tests and Inspections K. FIRE PROTECTION PROGRAM 1.0 Design Bases Summary 1.1 Defense-in-Depth 1.2 NFPA 805 Performance Criteria 1.3 Codes of Record 2.0 System Description 2.1 Required Systems 2.2 3.0 Safety Evaluation 4.0 Fire Protection Program Documentation, Configuration Control and Quality Assurance L. REMOTE SHUTDOWN SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections M. HYDROGEN WATER CHEMISTRY AND NOBLE METAL CHEMICAL ADDITION (NOBLECHEM) SYSTEMS 1.0 Design Basis 1.1 Noble Metal Chemical Addition System 1.2 Hydrogen Water Chemistry System NMP Unit 1 UFSAR Section Title TOC xx Rev. 25, October 2017 2.0 System Design 2.1 Noble Metal Chemical Addition 2.2 Hydrogen Water Chemistry System 2.2.1 HWC Feedwater Hydrogen Injection 2.2.2 HWC Offgas Oxygen Injection 2.2.3 HWC Offgas Sample 3.0 System Evaluation 4.0 Tests and Inspections N. REFERENCES APPENDIX 10A This Section Deleted APPENDIX 10B This Section Deleted SECTION XI STEAM-TO-POWER CONVERSION SYSTEM A. DESIGN BASES B. SYSTEM DESIGN AND OPERATION 1.0 Turbine Generator 2.0 Turbine Condenser 3.0 Condenser Air Removal and Offgas System 4.0 Circulating Water System 5.0 Condensate Pumps 6.0 Condensate Filtration System 6.0A Condensate Demineralizer System 7.0 Condensate Transfer System 8.0 Feedwater Booster Pumps 9.0 Feedwater Pumps 10.0 Feedwater Heaters C. SYSTEM ANALYSIS D. TESTS AND INSPECTIONS SECTION XII RADIOLOGICAL CONTROLS A. RADIOACTIVE WASTES 1.0 Design Bases 1.1 Objectives 1.2 Types of Radioactive Wastes 1.2.1 Gaseous Waste NMP Unit 1 UFSAR Section Title TOC xxi Rev. 25, October 2017 1.2.2 Liquid Wastes 1.2.3 Solid Wastes 2.0 System Design and Evaluation 2.1 Gaseous Waste System 2.1.1 Offgas System 2.1.2 Steam-Packing Exhausting System 2.1.3 Building Ventilation Systems 2.1.4 Stack 2.2 Liquid Waste System 2.2.1 Liquid Waste Handling Processes 2.2.2 Sampling and Monitoring Liquid Wastes 2.2.3 Liquid Waste Equipment Arrangement 2.2.4 Liquid Radioactive Waste System Control 2.3 Solid Waste System 2.3.1 Solid Waste Handling Processes 2.3.2 Solid Waste System Equipment 2.3.3 Process Control Program 3.0 Safety Limits 4.0 Tests and Inspections 4.1 Waste Process Systems 4.2 Filters 4.3 Effluent Monitors 4.3.1 Offgas and Stack Monitors 4.3.2 Liquid Waste Effluent Monitor B. RADIATION PROTECTION 1.0 Primary and Secondary Shielding 1.1 Design Bases 1.2 Design 1.2.1 Reactor Shield Wall 1.2.2 Biological Shield 1.2.3 Miscellaneous 1.3 Evaluation 2.0 Area Radioactivity Monitoring Systems 2.1 Area Radiation Monitoring System 2.1.1 Design Bases 2.1.2 Design 2.1.3 Evaluation 2.2 Area Air Contamination Monitoring System 2.2.1 Design Bases NMP Unit 1 UFSAR Section Title TOC xxii Rev. 25, October 2017 2.2.2 Design 2.2.3 Evaluation 3.0 Radiation Protection 3.1 Facilities 3.1.1 Laboratory, Counting Room and Calibration Facilities 3.1.2 Change Room and Shower Facilities 3.1.3 Personnel Decontamination Facility 3.1.4 Tool and Equipment Decontamination Facility 3.2 Radiation Control 3.2.1 Shielding 3.2.2 Access Control 3.3 Contamination Control 3.3.1 Facility Contamination Control 3.3.2 Personnel Contamination Control 3.3.3 Airborne Contamination Control 3.4 Personnel Dose Determinations 3.4.1 Radiation Dose 3.5 Radiation Protection Instrumentation 3.5.1 Counting Room Instrumentation 3.5.2 Portable Radiation Instrumentation 3.5.3 Air Sampling Instrumentation 3.5.4 Personnel Monitoring Instruments 3.5.5 Emergency Instrumentation 4.0 Tests and Inspections 4.1 Shielding 4.2 Area Radiation Monitors 4.3 Area Air Contamination Monitors 4.4 Radiation Protection Facilities 4.4.1 Ventilation Air Flows 4.4.2 Instrument Calibration Well Shielding 4.5 Radiation Protection Instrumentation SECTION XIII CONDUCT OF OPERATIONS A. ORGANIZATION AND RESPONSIBILITY 1.0 Offsite Organization 1.1 Station Organization 1.1.1 This SectionDeleted 1.1.2 This Section Deleted NMP Unit 1 UFSAR Section Title TOC xxiii Rev. 25, October 2017 1.1.3 This Section Deleted 2.0 Nine Mile Point Nuclear Station, LLC, Organization 2.1 This Section Deleted 2.2 This Section Deleted 2.3 This Section Deleted 3.0 This Section Deleted 4.0 Operating Shift Crews 5.0 Qualifications of Staff Personnel B. QUALIFICATIONS AND TRAINING OF PERSONNEL 1.0 This section Deleted 2.0 This section Deleted 3.0 This section Deleted 4.0 Training of Personnel 4.1 General Responsibility 4.2 This Section Deleted 4.3 This Section Deleted 4.3.1 This Section Deleted 4.3.2 This Section Deleted 4.3.3 This Section Deleted 4.3.4 This Section Deleted 4.3.5 This Section Deleted 4.3.6 This Section Deleted 4.3.7 This Section Deleted 4.3.8 For Operations Director and Operations Shift Superintendent 4.4 Training of Licensed Operator Candidates/Licensed NRC Operator Retraining 5.0 Cooperative Training With Local, State and Federal Officials C. OPERATING PROCEDURES D. EMERGENCY PLAN AND PROCEDURES E. SECURITY F. RECORDS 1.0 Operations 1.1 Control Room Log 1.2 Shift Manager's Log NMP Unit 1 UFSAR Section Title TOC xxiv Rev. 25, October 2017 1.3 Radwaste Log 1.4 Waste Quantity Level Shipped 2.0 Maintenance 3.0 Radiation Protection 3.1 Personnel Exposure 3.2 By-Product Material as Required by 10CFR30 3.3 Meter Calibrations 3.4 Station Radiological Conditions in Accessible Areas 3.5 Administration of the Radiation Protection Program and Procedures 4.0 Chemistry and Radiochemistry 5.0 Special Nuclear Materials 6.0 Calibration of Instruments 7.0 Administrative Records and Reports G. REVIEW AND AUDIT OF OPERATIONS 1.0 This Section Deleted 1.1 This Section Deleted 2.0 This Section Deleted 2.1 This Section Deleted 3.0 This Section Deleted SECTION XIV INITIAL TESTING AND OPERATIONS A. TESTS PRIOR TO INITIAL REACTOR FUELING B. INITIAL CRITICALITY AND POSTCRITICALITY TESTS 1.0 Initial Fuel Loading and Near-Zero Power Tests at Atmospheric Pressure 1.1 General Requirements 1.2 General Procedures 1.3 Core Loading and Critical Test Program 2.0 Heatup from Ambient to Rated Temperature 2.1 General 2.2 Tests Conducted 3.0 From Zero to 100 Percent Initial Reactor Rating 4.0 Full-Power Demonstration Run NMP Unit 1 UFSAR Section Title TOC xxv Rev. 25, October 2017 5.0 Comparison of Base Conditions 6.0 Additional Tests at Design Rating 7.0 Startup Report SECTION XV SAFETY ANALYSIS A. INTRODUCTION B. BOUNDARY PROTECTION SYSTEMS 1.0 Transients Considered 2.0 Methods and Assumptions 3.0 Transient Analysis 3.1 Turbine Trip Without Bypass 3.1.1 Objectives 3.1.2 Assumptions and Initial Conditions 3.1.3 Comments 3.1.4 Results 3.2 Loss of 100°F Feedwater Heating 3.2.1 Objective 3.2.2 Assumptions and Initial Conditions 3.2.3 Results 3.3 Feedwater Controller Failure- Maximum Demand 3.3.1 Objective 3.3.2 Assumptions and Initial Conditions 3.3.3 Comments 3.3.4 Results 3.4 Control Rod Withdrawal Error 3.4.1 Objective 3.4.2 Assumptions and Initial Conditions 3.4.3 Comments 3.4.4 Results 3.5 Main Steam Line Isolation Valve Closure (With Scram) 3.5.1 Objective 3.5.2 Assumptions and Initial Conditions 3.5.3 Comments 3.5.4 Results 3.6 Inadvertent Startup of Cold Recirculation Loop 3.6.1 Objective 3.6.2 Assumptions and Initial Conditions 3.6.3 Comment 3.6.4 Results NMP Unit 1 UFSAR Section Title TOC xxvi Rev. 25, October 2017 3.7 Recirculation Pump Trips 3.7.1 Objectives 3.7.2 Assumptions and Initial Conditions 3.7.3 Comments 3.7.4 Results 3.8 Recirculation Pump Stall 3.8.1 Objective 3.8.2 Assumptions and InitialConditions 3.8.3 Comments 3.8.4 Results 3.9 Recirculation Flow Controller Malfunction - Increase Flow 3.9.1 Objective 3.9.2 Assumptions and Initial Conditions 3.9.3 Comments 3.9.4 Results 3.10 Flow Controller Malfunction - Decrease Flow 3.10.1 Objective 3.10.2 Assumptions and Initial Conditions 3.10.3 Comments 3.10.4 Results 3.11 Inadvertent Actuation of One Solenoid Relief Valve 3.11.1 Objectives 3.11.2 Assumptions and Initial Conditions 3.11.3 Comments 3.11.4 Results 3.12 Safety Valve Actuation (Overpressurization Analysis) 3.12.1 Objectives 3.12.2 Assumptions and Initial Conditions 3.12.3 Comments 3.12.4 Results 3.13 Feedwater Controller Malfunction (Zero Demand) 3.13.1 Objective 3.13.2 Assumptions and Initial Conditions 3.13.3 Comments 3.13.4 Results 3.14 Turbine Trip with Partial Bypass (Low Power) 3.14.1 Objectives 3.14.2 Assumptions and Initial Conditions NMP Unit 1 UFSAR Section Title TOC xxvii Rev. 25, October 2017 3.14.3 Comments 3.14.4 Results 3.15 Turbine Trip with Partial Bypass (Full Power) 3.15.1 Objectives 3.15.2 Assumptions and Initial Conditions 3.15.3 Comments 3.15.4 Results 3.16 Inadvertent Actuation of One Bypass Valve 3.16.1 Objectives 3.16.2 Assumptions and Initial Conditions 3.16.3 Comments 3.16.4 Results 3.17 One Feedwater Pump Trip and Restart 3.17.1 Objective 3.17.2 Assumptions and Initial Conditions 3.17.3 Comments 3.17.4 Results 3.18 Loss of Main Condenser Vacuum 3.19 Loss of Electrical Load (Generator Trip) 3.19.1 Objectives 3.19.2 Assumptions and Initial Conditions 3.19.3 Comments 3.19.4 Results 3.20 Loss of Auxiliary Power 3.20.1 Objective 3.20.2 Assumptions and Initial Conditions 3.20.3 Comments 3.20.4 Results 3.21 Pressure Regulator Malfunction 3.21.1 Objective 3.21.2 Assumptions and Initial Conditions 3.21.3 Comments 3.21.4 Results 3.22 Instrument Air Failure 3.22.1 Objective 3.22.2 Assumptions and Initial Conditions 3.22.3 Comments 3.22.4 Results 3.23 Dc Power Interruptions 3.23.1 Objective 3.23.2 Assumptions and Initial Conditions NMP Unit 1 UFSAR Section Title TOC xxviii Rev. 25, October 2017 3.23.3 Comments 3.23.4 Results 3.24 Failure of One Diesel Generator to Start 3.24.1 Objective 3.24.2 Assumptions and Initial Conditions 3.24.3 Comments 3.24.4 Results 3.25 Power Bus Loss of Voltage 3.25.1 Objective 3.25.2 Assumptions and Initial Conditions 3.25.3 Comments 3.25.4 Results C. STANDBY SAFEGUARDS ANALYSIS 1.0 Main Steam Line Break Outside the Drywell 1.1 Identification of Causes 1.2 Accident Analysis 1.2.1 Valve Closure Initiation 1.2.2 Feedwater Flow 1.2.3 Core Shutdown 1.2.4 Mixture Level 1.2.5 Subcooled Liquid 1.2.6 System Pressure and Steam-Water Mass 1.2.7 Mixture Impact Forces 1.2.8 Core Internal Forces 1.3 Radiological Effects 1.3.1 Radioactivity Releases 1.3.2 Meteorology and Dose Rates 2.0 Loss-of-Coolant Accident 2.1 Introduction 2.2 Input to Analysis 2.2.1 Operational and ECCS Input Parameters 2.2.2 Single Failure Study on ECCS Manually-Controlled Electrically-Operated Valves 2.2.3 Single Failure Basis 2.2.4 Pipe Whip Basis 2.2.5 Degraded Voltage/LOCA 2.3 This section deleted 2.4 Appendix K LOCA Performance NMP Unit 1 UFSAR Section Title TOC xxix Rev. 25, October 2017 Analysis 2.4.1 Computer Codes 2.4.2 Description of Model Changes 2.4.3 Analysis Procedure 2.4.3.1 BWR/2 Generic Analysis 2.4.3.2 Unit 1-Specific Analysis Break Spectrum Evaluation 2.4.4 Analysis Results 3.0 Refueling Accident 3.1 Identification of Causes 3.2 Accident Analysis 3.3 Radiological Effects 3.3.1 Fission Product Releases 3.3.2 Meteorology and Dose Rates 4.0 Control Rod Drop Accident 4.1 Identification of Causes 4.2 Accident Analysis 4.3 Designed Safeguards 4.4 Procedural Safeguards 4.5 Radiological Effects 4.5.1 Fission Product Releases 4.5.2 Meteorology and Dose Rates 5.0 Containment Design Basis Accident 5.1 Original Recirculation Line Rupture Analysis - With Core Spray 5.1.1 Purpose 5.1.2 Analysis Method and Assumptions 5.1.3 Core Heat Buildup 5.1.4 Core Spray System 5.1.5 Containment Pressure Immediately Following Blowdown 5.1.6 Containment Spray 5.1.7 Blowdown Effects on Core Components 5.1.8 Radiological Effects 5.1.8.1 Fission Product Releases 5.1.8.2 Meteorology and Dose Rates 5.2 Original Containment Design Basis Accident Analysis - Without Core Spray 5.2.1 Purpose 5.2.2 Core Heatup 5.2.3 Containment Response 5.3 Design Basis Reconstitution NMP Unit 1 UFSAR Section Title TOC xxx Rev. 25, October 2017 Suppression Chamber Heatup Analysis 5.3.1 Introduction 5.3.2 Input to Analysis 5.3.3 DBR Suppression Chamber Heatup Analysis 5.3.3.1 Computer Codes 5.3.3.2 Analysis Methods 5.3.3.3 Analysis Results for Containment Spray Design Basis Assumptions 5.3.3.4 Analysis Results for EOP Operation Assumptions 5.3.4 Conclusions 6.0 New Fuel Bundle Loading Error Analysis 6.1 Identification of Causes 6.2 Accident Analysis 6.3 Safety Requirements 7.0 Meteorological Models Used in Accident Analyses 7.1 Introduction 7.2 Atmospheric Dispersion Factor Calculations 7.2.1 Offsite - EAB and LPZ 7.2.2 Control Room and Technical Support Center (Excluding MSLB) 7.2.3 Control Room - MSLB Puff Release 7.3 Summary of Results 7.4 Exfiltration 7.5 Secondary Containment Drawdown 7.5.1 Introduction 7.5.2 Analysis 7.5.3 Results D. REFERENCES SECTION XVI SPECIAL TOPICAL REPORTS A. REACTOR VESSEL 1.0 Applicability of Formal Codes and Pertinent Certifications 2.0 Design Analysis 2.1 Code Approval Analysis 2.2 Steady-State Analysis 2.2.1 Basis for Determining Stresses NMP Unit 1 UFSAR Section Title TOC xxxi Rev. 25, October 2017 2.3 Pipe Reaction Calculations 2.4 Earthquake Loading Criteria and Analysis 2.4.1 Seismic Analysis for Core Shroud Repair Modification 2.5 Reactor Vessel Support Stress Design Criteria and Analysis 2.6 Strain Safety Margin for Reactor Vessels 2.6.1 Introduction 2.6.2 Strain Margin 2.6.3 Failure Probability 2.6.4 Results of Probability Analysis 2.6.5 Conclusions 2.7 Components Required for Safe Reactor Shutdown 2.7.1 Design Basis Load Combinations 2.7.2 Expected Stress and Deformation 2.7.2.1 Recirculation Line Break 2.7.2.2 Steam Line Break 2.7.2.3 Earthquake Loadings 2.7.3 Stresses and Deformations at Which the Component is Unable to Function and Margin of Safety 2.7.3.1 Recirculation Line Break 2.7.3.2 Steam Line Break 2.8 Safety Margins Against Ductile Fracture 3.0 Inspection and Test Report Summary 3.1 Materials 3.2 Fabrication and Inspection 4.0 Surveillance Provisions 4.1 Coupon Surveillance Program 4.2 Periodic Inspection 5.0 Core Shroud Repair Design Description 5.1 Horizontal Weld Repair 5.2 Vertical Weld Repair B. PRESSURE SUPPRESSION CONTAINMENT 1.0 Applicability of Formal Codes and Pertinent Certifications 2.0 Design Analysis 2.1 Code Approval Calculations Under NMP Unit 1 UFSAR Section Title TOC xxxii Rev. 25, October 2017 Rated Conditions 2.2 Ultimate Capability Under Accident Conditions 2.3 Capability to Withstand Internal Missiles and Jet Forces 2.4 Flooding Capabilities of the Containment 2.5 Drywell Air Gap 2.5.1 Tests and Inspections 2.6 Reactor Shield Wall 2.7 Compatibility of Dynamic Deformations Occurring in the Drywell, Torus, and Connecting Vent Pipes 2.8 Containment Penetrations 2.8.1 Classification of Penetrations 2.8.2 Design Bases 2.8.3 Method of Stress Analysis 2.8.4 Leak Test Capability 2.8.5 Fatigue Design 2.8.6 Material Specification 2.8.7 Applicable Codes 2.8.8 Jet and Reaction Loads 2.9 Drywell Shear Resistance Capability and Support Skirt Junction Stresses 3.0 Inspection and Test Report Summary 3.1 Fabrication and Inspection 3.2 Tests Conducted 3.3 Discussion of Results 3.3.1 Results 3.3.2 Effect of Various Transients 3.3.2.1 Ambient Temperature and Solar Heating of Shell 3.3.2.2 Thermal Lag Through Reference Chamber Wall 3.3.2.3 Condensation in Reference Chamber 3.3.2.4 Volume Changes Due to Thermal Transients 3.3.2.5 Overpressure Test--Plate Stresses C. ENGINEERED SAFEGUARDS 1.0 Seismic Analysis and Stress Report 1.1 Introduction 1.2 Mathematical Model NMP Unit 1 UFSAR Section Title TOC xxxiii Rev. 25, October 2017 1.3 Method of Analysis 1.3.1 Flexibility or Influence Coefficient Matrix 1.3.2 Normal Mode Frequencies and Mode Shapes 1.3.3 The Seismic Spectrum Values 1.3.4 Dynamic Modal Loads 1.3.5 Modal Response Quantities 1.3.6 The Combined Response Quantities 1.3.7 Basic Criteria for Analysis 1.4 Discussion of Results 2.0 Containment Spray System 2.1 Design Adequacy at Rated Conditions 2.1.1 General 2.1.2 Condensation and Heat Removal Mechanisms 2.1.3 Mechanical Design 2.1.4 Loss-of-Coolant Accident 2.2 Summary of Test Results 2.2.1 Spray Tests Conducted 3.0 Core Spray and Containment Spray Suction Strainers D. DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 1.0 Classification and Seismic Criteria 1.1 Design Techniques 1.1.1 Structures 1.1.2 Systems and Components 1.2 Pipe Supports 1.3 Seismic Exposure Assumptions 2.0 Plant Design for Protection Against Postulated Piping Failures in High-Energy Lines 2.1 Inside Primary Containment 2.1.1 Containment Integrity Analysis 2.1.1.1 Fluid Forces 2.1.1.2 Impact Velocities and Effects 2.1.2 Systems Affected by Line Break 2.1.3 Engineered Safeguards Protection 2.2 Outside Primary Containment 3.0 Building Separation Analysis 4.0 Tornado Protection 5.0 Thermally-Induced Overpressurization NMP Unit 1 UFSAR Section Title TOC xxxiv Rev. 25, October 2017 of Isolated Piping E. EXHIBITS F. CONTAINMENT DESIGN REVIEW G. USAGE OF CODES/STANDARDS FOR STRUCTURAL STEEL AND CONCRETE H. REFERENCES SECTION XVII ORIGINAL ENVIRONMENTAL STUDIES A. METEOROLOGY 1.0 General 2.0 Synoptic Meteorological Factors 3.0 Micrometeorology 3.1 Wind Patterns 3.1.1 200-Ft Wind Roses 3.1.2 Estimates of Winds at the 350-Ft Level 3.1.3 Comparison Between Tower and Satellite Winds 3.2 Lapse Rate Distributions 3.3 Turbulence Classes 3.4 Dispersion Parameters 3.4.1 Changes in Dispersion Parameters 4.0 Applications to Release Problems 4.1 Concentrations From a Ground-Level Source 4.2 Concentrations From an Elevated Source 4.3 Radial Concentrations 4.3.1 Monthly and Annual Sector Concentrations 4.4 Least Favorable Concentrations Over an Extended Period 4.4.1 Ground-Level Release 4.4.2 Elevated Release 4.5 Mean Annual Sector Deposition 4.6 Dose Rates From a Plume of Gamma Emitters 4.6.1 RADOS Program 4.6.2 Centerline Dose Rates NMP Unit 1 UFSAR Section Title TOC xxxv Rev. 25, October 2017 4.6.3 Sector Dose Rates 4.7 Concentrations From a Major Steam Line Break 5.0 Conclusions B. LIMNOLOGY 1.0 Introduction 2.0 Summary Report of Cruises 3.0 Dilution of Station Effluent in Selected Areas 3.1 Dilution of Effluent at the Lake Surface Above the Discharge 3.2 Dilution of Effluent at the Site Boundaries 3.2.1 General 3.2.2 Dilution of Effluent at the Eastern Site Boundary 3.2.3 Dilution of Effluent West of the Station Site 3.3 Dilution of Effluent at the City of Oswego Intake 3.3.1 Tilting of the Isothermal Planes and Subsequent Dilution 3.3.2 Dilution as a Function of Current Velocity 3.3.3 Percent of Time Effluent Will be Carried to the Oswego Area 3.3.4 Mixing With Distance 3.3.5 Oswego River Water as a Buffer to Prevent Effluent From Passing Over the Intake 3.3.6 Summary of Annual Dilution Factors for the City of Oswego Intake 3.4 Dilution of Effluent at the Nine Mile Point Intake 3.5 Summary of Dilution in the Nine Mile Point Area 4.0 Preliminary Study of Lake Biota Off Nine Mile Point 4.1 Biological Studies 4.1.1 Plankton Study 4.1.2 Bottom Study 4.2 Summary of Biological Studies 5.0 Conclusions NMP Unit 1 UFSAR Section Title TOC xxxvi Rev. 25, October 2017 C. EARTH SCIENCES 1.0 Introduction 2.0 Additional Subsurface Studies 3.0 Construction Experience 3.1 Station Area 3.2 Intake and Discharge Tunnels 4.0 Correlation With Previous Studies 4.1 General 4.2 Geological Conditions 4.3 Hydrological Conditions 4.4 Seismological Conditions 4.5 Conclusion D. REFERENCES SECTION XVIII HUMAN FACTORS ENGINEERING/SAFETY PARAMETER DISPLAY SYSTEM A. DETAILED CONTROL ROOM DESIGN REVIEW 1.0 General 2.0 Planning Requirements for the DCRDR 3.0 DCRDR Review Process 3.1 Operator Survey 3.2 Historical Review 3.3 Task Analysis 3.4 Control Room Inventory 3.5 Control Room Survey 3.6 Verification of Task Performance Capabilities 3.7 Validation of Control Room -4 3.8 Compilation of Discrepancy Findings 4.0 Assessment and Implementation 4.1 Assessment 4.2 Implementation 4.2.1 Integrated Cosmetic Package 4.2.2 Functional Fixes 5.0 Reporting 6.0 Continuing Human Factors Program 6.1 Fix Verifications 6.2 Multidisciplinary Review Team Assessments 6.3 Human Factors Manual for Future Design Change NMP Unit 1 UFSAR Section Title TOC xxxvii Rev. 25, October 2017 6.4 Outstanding Human Factors Items 7.0 References B. SAFETY PARAMETER DISPLAY SYSTEM 1.0 Introduction to the Safety Parameter Display System 2.0 System Description 3.0 Role of the SPDS 4.0 Human Factors Engineering Guidelines 5.0 Human Factors Engineering Principles Applied to the SPDS Design 5.1 NUREG-0737, Supplement 1, Section 4.1.a 5.1.1 Concise Display 5.1.2 Critical Plant Variables 5.1.3 Rapid and Reliable Determination of Safety Status 5.1.4 Aid to Control Room Personnel 5.2 NUREG-0737, Supplement 1, Section 4.1.b 5.2.1 Convenient Location 5.2.2 Continuous Display 5.3 NUREG-0737, Supplement 1, Section 4.1.c 5.3.1 Procedures and Training 5.3.2 Isolation of SPDS from Safety-Related Systems 5.4 NUREG-0737, Supplement 1, Section 4.1.e 5.4.1 Incorporation of Accepted Human Factors Engineering Principles 5.4.2 Information Can Be Readily Perceived and Comprehended 5.5 NUREG-0737, Supplement 1, Section 4.1.f, Sufficient Information 6.0 Procedures 6.1 Operating Procedures 6.2 Surveillance Procedures 7.0 References APPENDIX A Unused NMP Unit 1 UFSAR Section Title TOC xxxviii Rev. 25, October 2017 APPENDIX B NINE MILE POINT NUCLEAR STATION, LLC, QUALITY ASSURANCE PROGRAM TOPICAL REPORT, NINE MILE POINT NUCLEAR STATION UNITS 1 AND 2 OPERATIONS PHASE APPENDIX C LICENSE RENEWAL SUPPLEMENT - AGING MANAGEMENT PROGRAMS AND TIME-LIMITED AGING ANALYSES NMP Unit 1 UFSAR LIST OF TABLES Table Number Title TOC xxxix Rev. 25, October 2017 I-1 COMPARISON TO STANDARDS - HISTORICAL (PROVIDED WITH APPLICATION TO CONVERT TO FULL-TERM OPERATING LICENSE) I-2 ABBREVIATIONS AND ACRONYMS USED IN UFSAR II-1 1980 POPULATION AND POPULATION DENSITY FOR TOWNS AND CITIES WITHIN 12 MILES OF NINE MILE POINT - UNIT 1 II-2 CITIES WITHIN A 50-MILE RADIUS OF THE STATION WITH POPULATIONS OVER 10,000 II-3 REGIONAL AGRICULTURAL USE II-4 REGIONAL AGRICULTURAL STATISTICS - CATTLE AND MILK PRODUCTION II-5 INDUSTRIAL FIRMS WITHIN 8 KM (5 MI) OF UNIT 1 II-6 PUBLIC UTILITIES IN OSWEGO COUNTY II-7 PUBLIC WATER SUPPLY DATA FOR LOCATIONS WITHIN AN APPROXIMATE 30-MILE RADIUS II-8 RECREATIONAL AREAS IN THE REGION II-9 SOURCES OF TOXIC CHEMICALS WITHIN 8 KM (5 MI) OF UNIT 1 SITE II-10 PREDICTED VAPOR CONCENTRATION IN THE UNIT 1 CONTROL ROOM V-1 REACTOR COOLANT SYSTEM DATA V-2 OPERATING CYCLES AND TRANSIENT ANALYSIS RESULTS V-3 FATIGUE RESISTANCE ANALYSIS V-4 CODES FOR SYSTEMS FROM REACTOR VESSEL CONNECTION TO SECOND ISOLATION VALVE V-5 TIME TO AUTOMATIC BLOWDOWN NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xl Rev. 25, October 2017 V-6 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VI-1 DRYWELL PENETRATIONS VI-2 SUPPRESSION CHAMBER PENETRATIONS VI-3a REACTOR COOLANT SYSTEM ISOLATION VALVES VI-3b PRIMARY CONTAINMENT ISOLATION VALVES - LINES ENTERING FREE SPACE OF THE CONTAINMENT VI-4 SEISMIC DESIGN CRITERIA FOR ISOLATION VALVES VI-5 INITIAL TESTS PRIOR TO STATION OPERATION VII-1 PERFORMANCE TESTS VIII-1 ASSOCIATION BETWEEN PRIMARY SAFETY FUNCTIONS AND EMERGENCY OPERATING PROCEDURES VIII-2 LIST OF EOP KEY PARAMETERS VIII-3 TYPE AND INSTRUMENT CATEGORY FOR UNIT 1 RG 1.97 VARIABLES VIII-4 PROTECTIVE SYSTEM FUNCTION VIII-5 NON-TECHNICAL SPECIFICATION INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK IX-1 MAGNITUDE AND DUTY CYCLE OF MAJOR STATION BATTERY LOADS XII-1 FLOWS AND ACTIVITIES OF MAJOR SOURCES OF GASEOUS ACTIVITY XII-2 QUANTITIES AND ACTIVITIES OF LIQUID RADIOACTIVE WASTES XII-3 ANNUAL SOLID WASTE ACCUMULATION AND ACTIVITY LIST OF TABLES (Cont'd.) NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xli Rev. 25, October 2017 XII-4 LIQUID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS XII-5 SOLID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS XII-6 OCCUPANCY TIMES XII-7 GAMMA ENERGY GROUPS XII-8 AREA RADIATION MONITOR DETECTOR LOCATIONS XIII-1 TABLE DELETED XIII-2 TABLE DELETED XV-1 TABLE DELETED XV-2 TRIP POINTS FOR PROTECTIVE FUNCTIONS XV-3 thru TABLES DELETED XV-4 XV-5 BLOWDOWN RATES XV-6 REACTOR COOLANT CONCENTRATIONS (µCi/gm) XV-7 TABLE DELETED XV-7a MSLB ACCIDENT ANALYSIS INPUTS AND ASSUMPTIONS XV-7b MSLB ACCIDENT RELEASE RATES XV-8 MAIN STEAM LINE BREAK ACCIDENT DOSES XV-9 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS XV-9a TABLE DELETED XV-9b CORE SPRAY SYSTEM FLOW PERFORMANCE ASSUMED IN THE LOCA ANALYSIS XV-10 ECCS SINGLE VALVE FAILURE ANALYSIS NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xlii Rev. 25, October 2017 XV-11 SINGLE FAILURES CONSIDERED IN LOCA ANALYSIS XV-12 TRACG-LOCA Licensing Results for Nine Mile Point 1 XV-13 thru TABLES DELETED XV-21a XV-22 ACTIVITY RELEASED TO THE REACTOR BUILDING FOLLOWING THE FHA (CURIES) XV-23 UNIFORM UNFILTERED STACK DISCHARGE RATES FROM 0 TO 2 HR AFTER THE FHA (CURIES/SEC) XV-24 FUEL HANDLING ACCIDENT DOSES XV-25 FHA ANALYSIS INPUTS AND ASSUMPTIONS XV-26 CRD ACCIDENT ANALYSIS INPUTS AND ASSUMPTIONS XV-27 CRDA NOBLE GAS RELEASE XV-28 CRDA HALOGEN RELEASE XV-29 CONTROL ROD DROP ACCIDENT DOSES XV-29a WETTING OF FUEL CLADDING BY CORE SPRAY XV-29b POST-LOCA AIRBORNE DRYWELL FISSION PRODUCT INVENTORY (CURIES) XV-29c POST-LOCA REACTOR BUILDING FISSION PRODUCT INVENTORY (CURIES) XV-29d POST-LOCA DISCHARGE RATES (CURIES/SEC) XV-30 CORE FISSION PRODUCT INVENTORY XV-31 LOCA ANALYSIS INPUTS AND ASSUMPTIONS XV-32 LOSS-OF-COOLANT ACCIDENT DOSES NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xliii Rev. 25, October 2017 XV-32a SIGNIFICANT INPUT PARAMETERS TO THE DBR CONTAINMENT SUPPRESSION CHAMBER HEATUP ANALYSIS XV-33 TABLE DELETED XV-34 TABLE DELETED XV-34a RELEASE/INTAKE ELEVATIONS XV-34b RELEASE/INTAKE DISTANCE AND DIRECTIONS XV-35 TABLE DELETED XV-35a X/Q VALUES FOR THE CONTROL ROOM XV-35b X/Q VALUES FOR THE TECHNICAL SUPPORT CENTER XV-35c OFFSITE X/Q VALUES FOR GROUND-LEVEL RELEASES XV-35d OFFSITE X/Q VALUES FOR ELEVATED RELEASES XV-36 REACTOR BUILDING LEAKAGE PATHS XVI-1 CODE CALCULATION SUMMARY XVI-2 STEADY-STATE - (100% FULL POWER NORMAL OPERATION) PERTINENT STRESSES OR STRESS INTENSITIES XVI-3 LIST OF REACTIONS FOR REACTOR VESSEL NOZZLES XVI-4 EFFECT OF VALUE OF INITIAL FAILURE PROBABILITY XVI-5 SINGLE TRANSIENT EVENT FOR REACTOR PRESSURE VESSEL XVI-6 POSTULATED EVENTS XVI-7 MAXIMUM STRAINS FROM POSTULATED EVENTS XVI-8 CORE STRUCTURE ANALYSIS RECIRCULATION LINE BREAK XVI-9 CORE STRUCTURE ANALYSIS STEAM LINE BREAK NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xliv Rev. 25, October 2017 XVI-9a CORE SHROUD REPAIR DESIGN SUPPORTING DOCUMENTATION XVI-10 DRYWELL JET AND MISSILE HAZARD ANALYSIS DATA XVI-11 DRYWELL JET AND MISSILE HAZARD ANALYSIS RESULTS XVI-12 STRESS DUE TO DRYWELL FLOODING XVI-13 ALLOWABLE WELD SHEAR STRESS XVI-14 LEAK RATE TEST RESULTS XVI-15 OVERPRESSURE TEST--PLATE STRESSES XVI-16 STRESS SUMMARY XVI-17 HEAT TRANSFER COEFFICIENTS AS A FUNCTION OF DROP DIAMETER XVI-18 HEAT TRANSFER COEFFICIENT AS A FUNCTION OF PRESSURE XVI-19 RELATIONSHIP BETWEEN PARTICLE SIZE AND TYPE OF SPRAY PATTERN XVI-20 ALLOWABLE STRESSES FOR FLOOR SLABS, BEAMS, COLUMNS, WALLS, FOUNDATIONS, ETC. XVI-21 ALLOWABLE STRESSES FOR STRUCTURAL STEEL XVI-22 ALLOWABLE STRESSES - REACTOR VESSEL CONCRETE PEDESTAL XVI-23 DRYWELL - ANALYZED DESIGN LOAD COMBINATIONS XVI-24 SUPPRESSION CHAMBER - ANALYZED DESIGN LOAD COMBINATIONS XVI-25 ACI CODE 505 ALLOWABLE STRESSES AND ACTUAL STRESSES FOR CONCRETE VENTILATION STACK NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xlv Rev. 25, October 2017 XVI-26 ALLOWABLE STRESSES FOR CONCRETE SLABS, WALLS, BEAMS, STRUCTURAL STEEL, AND CONCRETE BLOCK WALLS XVI-27 SYSTEM LOAD COMBINATIONS XVI-28 HIGH-ENERGY SYSTEMS - INSIDE CONTAINMENT XVI-29 HIGH-ENERGY SYSTEMS - OUTSIDE CONTAINMENT XVI-30 SYSTEMS WHICH MAY BE AFFECTED BY PIPE WHIP XVI-31 CAPABILITY TO RESIST WIND PRESSURE AND WIND VELOCITY XVII-1 DISPERSION AND ASSOCIATED METEOROLOGICAL PARAMETERS XVII-2 RELATION OF SATELLITE AND NINE MILE POINT WINDS XVII-3 FREQUENCY OF OCCURRENCE OF LAPSE RATES - 1963 AND 1964 XVII-4 RELATION BETWEEN WIND DIRECTION RANGE AND TURBULENCE CLASSES XVII-5 STACK CHARACTERISTICS XVII-6 DISTRIBUTION OF TURBULENCE CLASSES BY SECTORS XVII-7 SECTOR CONCENTRATIONS - 1963 SECTOR A SOURCE XVII-8 SECTOR CONCENTRATIONS - 1963 SECTOR B SOURCE XVII-9 SECTOR CONCENTRATIONS - 1963 SECTOR C SOURCE XVII-10 SECTOR CONCENTRATIONS - 1963 SECTOR D1 SOURCE XVII-11 SECTOR CONCENTRATIONS - 1963 SECTOR D2 SOURCE NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xlvi Rev. 25, October 2017 XVII-12 SECTOR CONCENTRATIONS - 1963 SECTOR E SOURCE XVII-13 SECTOR CONCENTRATIONS - 1963 SECTOR F SOURCE XVII-14 SECTOR CONCENTRATIONS - 1963 SECTOR G SOURCE XVII-15 SECTOR CONCENTRATIONS - 1963 SECTOR A SOURCE HEIGHT GROUND XVII-16 SECTOR CONCENTRATIONS - 1963 SECTOR B SOURCE HEIGHT GROUND XVII-17 SECTOR CONCENTRATIONS - 1963 SECTOR C SOURCE HEIGHT GROUND XVII-18 SECTOR CONCENTRATIONS - 1963 SECTOR D1 SOURCE HEIGHT GROUND XVII-19 SECTOR CONCENTRATIONS - 1963 SECTOR D2 SOURCE HEIGHT GROUND XVII-20 SECTOR CONCENTRATIONS - 1963 SECTOR E SOURCE HEIGHT GROUND XVII-21 SECTOR CONCENTRATIONS - 1963 SECTOR F SOURCE HEIGHT GROUND XVII-22 SECTOR CONCENTRATIONS - 1963 SECTOR G SOURCE HEIGHT GROUND XVII-23 ESTIMATES OF THE LEAST FAVORABLE 30 DAYS IN 100 YEARS XVII-24 CONCENTRATIONS IN THE LEAST FAVORABLE CALENDAR MONTH - 1963-64 XVII-25 ANNUAL AVERAGE SECTOR DEPOSITION RATES (Vg = 0.5 CM/SEC) NMP Unit 1 UFSAR LIST OF TABLES (Cont.d) Tables Number Title TOC xlvii Rev. 25, October 2017 XVII-26 ANNUAL AVERAGE SECTOR DEPOSITION RATES (Vg = 2.5 CM/SEC) XVII-27 PRINCIPAL RADIONUCLIDES IN GASEOUS WASTE RELEASE XVII-28 CORRECTION FACTORS TO OBTAIN ADJUSTED CENTERLINE DOSE RATES FOR SECTOR ESTIMATES XVII-29 ANNUAL AVERAGE GAMMA DOSE RATES XVII-30 DILUTION CALCULATION FOR EASTWARD CURRENTS BASED ON WATER AVAILABILITY XVIII-1 SPDS PARAMETER SET NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title TOC xlviii Rev. 25, October 2017 I-1 PIPING, INSTRUMENT AND EQUIPMENT SYMBOLS II-1 STATION LOCATION II-2 AREA MAP II-3 SITE TOPOGRAPHY II-4 POPULATION DISTRIBUTION WITHIN A 12 MILE RADIUS OF THE STATION II-5 COUNTIES AND TOWNS WITHIN 12 MILES OF THE STATION II-6 1980 POPULATION DISTRIBUTION WITHIN A 50 MILE RADIUS OF THE STATION III-1 PLOT PLAN III-2 STATION FLOOR PLAN - ELEVATION 225-6 III-3 STATION FLOOR PLAN - ELEVATIONS 237-0 AND 250-0 III-4 STATION FLOOR PLAN - ELEVATION 261-0 III-5 STATION FLOOR PLAN - ELEVATIONS 277-0 AND 281-0 III-6 STATION FLOOR PLAN - ELEVATIONS 281-0 AND 291-0 III-7 STATION FLOOR PLAN - ELEVATIONS 298-0 AND 300-0 III-8 STATION FLOOR PLAN - ELEVATIONS 317-6 AND 318-0 III-9 STATION FLOOR PLAN - ELEVATIONS 320-0, 333-8, 340-0 AND 369-0 III-10 SECTION BETWEEN COLUMN ROWS 7 AND 8 III-11 SECTION BETWEEN COLUMN ROWS 12 AND 14 III-12 TURBINE BUILDING VENTILATION SYSTEM NMP Unit 1 UFSAR Figure Number Title TOC xlix Rev. 25, October 2017 III-13 LABORATORY AND RADIATION PROTECTION FACILITY VENTILATION SYSTEM III-14 CONTROL ROOM VENTILATION SYSTEM III-15 WASTE DISPOSAL BUILDING VENTILATION SYSTEM III-16 WASTE DISPOSAL BUILDING EXTENSION VENTILATION SYSTEM III-17 OFFGAS BUILDING VENTILATION SYSTEM III-18 TECHNICAL SUPPORT CENTER VENTILATION SYSTEM III-19 CIRCULATING WATER CHANNELS UNDER SCREEN AND PUMP HOUSE - NORMAL OPERATION III-20 CIRCULATING WATER CHANNELS UNDER SCREEN AND PUMP HOUSE - SPECIAL OPERATIONS III-21 INTAKE AND DISCHARGE TUNNELS PLAN AND PROFILE III-22 STACK - PLAN AND ELEVATION III-23 STACK FAILURE - CRITICAL DIRECTIONS IV-1 LIMITING POWER/FLOW LINE (TYPICAL) IV-2 thru FIGURES DELETED IV-3 IV-4 TYPICAL CONTROL ROD - ISOMETRIC IV-5 FIGURE DELETED IV-6 CONTROL ROD DRIVE AND HYDRAULIC SYSTEM IV-7 CONTROL ROD DRIVE ASSEMBLY IV-8 TYPICAL CONTROL ROD TO DRIVE COUPLING - ISOMETRIC IV-9 REACTOR VESSEL ISOMETRIC NMP Unit 1 UFSAR Figure Number Title TOC l Rev. 25, October 2017 V-1 REACTOR EMERGENCY COOLANT SYSTEM V-2 REACTOR VESSEL NOZZLE LOCATION V-3 REACTOR VESSEL SUPPORT V-4 thru FIGURES DELETED V-7 V-8 EMERGENCY CONDENSER SUPPLY ISOLATION VALVES (TYPICAL OF 2) VI-1 DRYWELL AND SUPPRESSION CHAMBER VI-2 ELECTRICAL PENETRATIONS - HIGH VOLTAGE VI-3 ELECTRICAL PENETRATIONS - LOW VOLTAGE VI-4 PIPE PENETRATIONS - HOT VI-4a CLAMSHELL EXPANSION JOINT VI-5 TYPICAL PENETRATION FOR INSTRUMENT LINES VI-6 REACTOR BUILDING DYNAMIC ANALYSIS - ACCELERATION EAST-WEST DIRECTION VI-7 REACTOR BUILDING DYNAMIC ANALYSIS - DEFLECTIONS EAST-WEST DIRECTION VI-8 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING SHEAR EAST-WEST DIRECTION VI-9 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING MOMENT EAST-WEST DIRECTION VI-10 REACTOR BUILDING DYNAMIC ANALYSIS - ACCELERATION NORTH-SOUTH DIRECTION VI-11 REACTOR BUILDING DYNAMIC ANALYSIS - DEFLECTIONS NORTH-SOUTH DIRECTION NMP Unit 1 UFSAR Figure Number Title TOC li Rev. 25, October 2017 VI-12 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING SHEAR - NORTH-SOUTH DIRECTION VI-13 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING MOMENT - NORTH-SOUTH DIRECTION VI-14 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. ACCELERATION VI-15 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. DEFLECTION VI-16 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. SHEAR VI-17 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. MOMENT VI-18 TYPICAL DOOR SEALS VI-19 DETAILS OF REACTOR BUILDING AIR LOCKS VI-20 INSTRUMENT LINE ISOLATION VALVE ARRANGEMENT VI-21 TYPICAL FLOW CHECK VALVE VI-22 ISOLATION VALVE SYSTEM VI-23 DRYWELL COOLING SYSTEM VI-24 REACTOR BUILDING VENTILATION SYSTEM VII-1 CORE SPRAY SYSTEM VII-2 FIGURE DELETED VII-3 CONTAINMENT SPRAY SYSTEM VII-4 thru FIGURES DELETED VII-5 VII-6 LIQUID POISON SYSTEM NMP Unit 1 UFSAR Figure Number Title TOC lii Rev. 25, October 2017 VII-7 MINIMUM ALLOWABLE SOLUTION TEMPERATURE VII-8 FIGURE DELETED VII-9 TYPICAL CONTROL ROD VELOCITY LIMITER VII-10 CONTROL ROD HOUSING SUPPORT VII-11 HYDROGEN FLAMMABILITY LIMITS VII-12 COMBUSTIBLE GAS CONTROL SYSTEM VII-13 H2-O2 SAMPLING SYSTEM VII-14 CONTROLLED HYDROGEN AND OXYGEN CONCENTRATIONS - INERTED CONTAINMENT VII-15 INTEGRATED POST-LOCA NITROGEN VOLUME REQUIREMENT - INERTED CONTAINMENT VII-16 CONTAINMENT PRESSURE FOLLOWING CAD ACTUATION - INERTED CONTAINMENT VII-17 FEEDWATER DELIVERY CAPABILITY (SHAFT DRIVEN PUMP) TO TIME AFTER TURBINE TRIP FOR 1000 PSIG REACTOR PRESSURE AND 1.0 INCH HG ABS EXHAUST PRESSURE VIII-1 FIGURE DELETED VIII-2 REACTOR PROTECTION SYSTEM ELEMENTARY DIAGRAM VIII-3 PROTECTIVE SYSTEM TYPICAL SENSOR ARRANGEMENT VIII-4 RECIRCULATION FLOW AND TURBINE CONTROL VIII-5 NEUTRON MONITORING INSTRUMENT RANGES VIII-6 SOURCE RANGE MONITOR (SRM) VIII-7 SRM DETECTOR LOCATION VIII-8 INTERMEDIATE RANGE MONITOR (IRM) NMP Unit 1 UFSAR Figure Number Title TOC liii Rev. 25, October 2017 VIII-9 IRM CORE LOCATION VIII-10 LPRM LOCATION WITHIN CORE LATTICE VIII-11 LPRM AND APRM CORE LOCATION VIII-12 LOCAL POWER RANGE MONITOR (LPRM) AND AVERAGE POWER RANGE MONITORS (APRM) VIII-13 APRM SYSTEM - TYPICAL VIII-14 TRIP LOGIC FOR APRM SCRAM AND ROD BLOCK VIII-15 TRAVERSING IN-CORE PROBE VIII-16 ROD PATTERN DURING STARTUP VIII-17 RADIAL POWER DISTRIBUTION FOR CONTROL ROD PATTERN SHOWN IN FIGURE VIII-16 VIII-18 DISTANCE FROM WORST CONTROL ROD TO NEAREST ACTIVE IRM MONITOR VIII-19 MEASURED RESPONSE TIME OF INTERMEDIATE RANGE SAFETY INSTRUMENTATION VIII-20 ENVELOPE OF MAXIMUM APRM DEVIATION BY FLOW CONTROL REDUCTION IN POWER VIII-21 ENVELOPE OF MAXIMUM APRM DEVIATION FOR APRM TRACKING WITH ON UNITS CONTROL ROD WITHDRAWAL VIII-22 MAIN STEAM LINE RADIATION MONITOR VIII-23 REACTOR BUILDING VENTILATION RADIATION MONITOR VIII-24 OFFGAS SYSTEM RADIATION MONITOR VIII-25 EMERGENCY CONDENSER VENT RADIATION MONITOR VIII-26 STACK EFFLUENT AND LIQUID EFFLUENT RADIATION MONITORS NMP Unit 1 UFSAR Figure Number Title TOC liv Rev. 25, October 2017 VIII-26a STACK EFFLUENT AND LIQUID EFFLUENT RADIATION MONITORS VIII-26b SERVICE WATER DISCHARGE LIQUID EFFLUENT RADIATION MONITOR VIII-27 CONTAINMENT SPRAY HEAT EXCHANGER RAW WATER EFFLUENT RADIATION MONITOR VIII-28 CONTAINMENT ATMOSPHERIC MONITORING SYSTEM VIII-29 ROD WORTH MINIMIZER IX-1 A.C. STATION POWER DISTRIBUTION IX-2 CONTROL AND INSTRUMENT POWER IX-3 TRAYS BELOW ELEVATION 261 IX-4 TRAYS BELOW ELEVATION 277 IX-5 TRAYS BELOW ELEVATION 300 IX-6 FIGURE DELETED IX-7 DIESEL GENERATOR LOADING FOR ORDERLY SHUTDOWN IX-8 DIESEL GENERATOR FUEL OIL SYSTEM X-1 REACTOR SHUTDOWN COOLING SYSTEM X-2 REACTOR CLEANUP SYSTEM X-3 CONTROL ROD DRIVE HYDRAULIC SYSTEM X-4 REACTOR BUILDING CLOSED LOOP COOLING SYSTEM X-5 TURBINE BUILDING CLOSED LOOP COOLING SYSTEM X-6 SERVICE WATER SYSTEM X-7 FIGURE DELETED NMP Unit 1 UFSAR Figure Number Title TOC lv Rev. 25, October 2017 X-8 SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM X-9 BREATHING, INSTRUMENT, AND SERVICE AIR X-10 REACTOR REFUELING SYSTEM PICTORIAL X-11 CASK DROP PROTECTION SYSTEM XI-1 STEAM FLOW AND REHEATER VENTILATION SYSTEM XI-2 EXTRACTION STEAM FLOW XI-3 MAIN CONDENSER AIR REMOVAL AND OFFGAS SYSTEM XI-4 CIRCULATING WATER SYSTEM XI-5 CONDENSATE FLOW XI-6 CONDENSATE TRANSFER SYSTEM XI-7 FEEDWATER FLOW SYSTEM XII-1 RADIOACTIVE WASTE DISPOSAL SYSTEM XIII-1 thru FIGURES DELETED XIII-5 XV-1 STATION TRANSIENT DIAGRAM XV-2 FIGURE DELETED XV-3 PLANT RESPONSE TO LOSS OF 100°F FEEDWATER HEATING XV-4 thru FIGURES DELETED XV-7 XV-8 STARTUP OF COLD RECIRCULATION LOOP - PARTIAL POWER XV-9 RECIRCULATION PUMP TRIPS (1 PUMP) XV-10 RECIRCULATION PUMP TRIPS (5 PUMPS) NMP Unit 1 UFSAR Figure Number Title TOC lvi Rev. 25, October 2017 XV-11 RECIRCULATION PUMP STALL XV-12 FLOW CONTROLLER MALFUNCTION (INCREASED FLOW) XV-13 FLOW CONTROLLER MALFUNCTION DECREASING FLOW XV-14 INADVERTENT ACTUATION OF ONE SOLENOID RELIEF VALVE XV-15 thru FIGURES DELETED XV-16 XV-17 FEEDWATER CONTROLLER MALFUNCTION - ZERO FLOW XV-18 TURBINE TRIP WITH PARTIAL BYPASS INTERMEDIATE POWER XV-19 TURBINE TRIP WITH PARTIAL BYPASS XV-20 INADVERTENT ACTUATION OF ONE BYPASS VALVE XV-21 ONE FEEDWATER PUMP TRIP AND RESTART XV-22 LOSS OF ELECTRICAL LOAD XV-23 LOSS OF AUXILIARY POWER XV-24 PRESSURE REGULATOR MALFUNCTION XV-25 MAIN STEAM LINE BREAK - COOLANT LOSS XV-26 thru FIGURES DELETED XV-56 XV-56A thru FIGURES DELETED XV-56C XV-56D LOSS-OF-COOLANT ACCIDENT - WITH CORE SPRAY CLADDING TEMPERATURE XV-56E LOSS-OF-COOLANT ACCIDENT DRYWELL PRESSURE NMP Unit 1 UFSAR Figure Number Title TOC lvii Rev. 25, October 2017 XV-56F LOSS-OF-COOLANT ACCIDENT SUPPRESSION CHAMBER PRESSURE XV-56G LOSS-OF-COOLANT ACCIDENT CONTAINMENT TEMPERATURE - WITH CORE SPRAY XV-57 CONTAINMENT DESIGN BASIS CLAD TEMPERATURE RESPONSE - WITHOUT CORE SPRAY XV-58 CONTAINMENT DESIGN BASIS METAL-WATER REACTION XV-59 CONTAINMENT DESIGN BASIS CLAD PERFORATION WITHOUT CORE SPRAY XV-60 CONTAINMENT DESIGN BASIS CONTAINMENT TEMPERATURE - WITHOUT CORE SPRAY XV-60A DBR ANALYSIS SUPPRESSION POOL AND WETWELL AIRSPACE TEMPERATURE RESPONSE - CONTAINMENT SPRAY DESIGN BASIS ASSUMPTION XV-60B DBR ANALYSIS SUPPRESSION POOL AND WETWELL AIRSPACE TEMPERATURE RESPONSE - EOP OPERATION ASSUMPTIONS XV-61 REACTOR BUILDING MODEL XV-62 EXFILTRATION VS. WIND SPEED - NORTHERLY WIND XV-63 REACTOR BUILDING DIFFERENTIAL PRESSURE XV-64 EXFILTRATION VS. WIND SPEED - SOUTHERLY WIND XV-65 REACTOR BUILDING - ISOMETRIC XV-66 REACTOR BUILDING - CORNER SECTIONS XV-67 REACTOR BUILDING - ROOF SECTIONS XV-68 REACTOR BUILDING - PANEL TO CONCRETE SECTIONS XV-69 REACTOR BUILDING - EXPANSION JOINT SECTIONS NMP Unit 1 UFSAR Figure Number Title TOC lviii Rev. 25, October 2017 XV-70 REACTOR BUILDING EXFILTRATION - NORTHERLY WIND XV-71 REACTOR BUILDING EXFILTRATION - SOUTHERLY WIND XV-72 REACTOR BUILDING DIFFERENTIAL PRESSURE XV-73 REACTOR BUILDING PRESSURE VS. TIME BY REACTOR BUILDING ELEVATION XV-74 REACTOR BUILDING PRESSURE VS. TIME BY REACTOR BUILDING ELEVATION (FOCUSED ON THE INITIAL 2.5 HR) XVI-1 SEISMIC ANALYSIS OF REACTOR VESSEL GEOMETRIC AND LUMPED MASS REPRESENTATION XVI-2 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. MOMENT XVI-3 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. SHEAR XVI-4 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. DEFLECTION XVI-5 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. ACCELERATION XVI-6 thru FIGURES DELETED XVI-8 XVI-9 REACTOR VESSEL SUPPORT STRUCTURE STRESS SUMMARY XVI-10 THERMAL ANALYSIS XVI-11 FAILURE PROBABILITY DENSITY FUNCTION XVI-12 ADDITION STRAINS PAST 4% REQUIRED TO EXCEED DEFINED SAFETY MARGIN XVI-12a SHROUD WELDS XVI-12b CORE SHROUD STABILIZERS NMP Unit 1 UFSAR Figure Number Title TOC lix Rev. 25, October 2017 XVI-12c CORE SHROUD WELDS XVI-12d V9/V10 VERTICAL WELD CLAMP ASSEMBLY XVI-13 LOSS-OF-COOLANT ACCIDENT - CONTAINMENT PRESSURE NO CORE OR CONTAINMENT SPRAYS XVI-14 FIGURE DELETED XVI-15 DRYWELL TO CONCRETE AIR GAP XVI-16 TYPICAL PENETRATIONS XVI-17 REACTOR SHIELD WALL CONSTRUCTION DETAILS XVI-18 VENT PIPE AND SUPPRESSION CHAMBER XVI-19 PRIMARY CONTAINMENT SUPPORT AND ANCHORAGE XVI-20 SEAL DETAILS - DRYWELL SHELL STEEL AND ADJACENT CONCRETE XVI-21 DRYWELL SLIDING - ACCELERATION, SHEAR, AND MOMENT XVI-22 SHEAR RESISTANCE CAPABILITY - INSIDE DRYWELL XVI-23 SHEAR RESISTANCE CAPABILITY - OUTSIDE DRYWELL XVI-24 DRYWELL - SUPPORT SKIRT JUNCTION STRESSES XVI-25 POINT LOCATION FOR CONTAINMENT SPRAY SYSTEM PIPING HEAT EXCHANGER TO DRYWELL XVI-26 COMPARISON OF STATIC AND DYNAMIC STRESSES (PSI) SEISMIC CONDITIONS - CONTAINMENT SPRAY SYSTEM HEAT EXCHANGER TO DRYWELL XVI-27 CONDUCTION IN A DROPLET XVI-28 LOSS OF COOLANT ACCIDENT - CONTAINMENT PRESSURE XVI-29 LOSS OF COOLANT ACCIDENT - CONTAINMENT PRESSURE NMP Unit 1 UFSAR Figure Number Title TOC lx Rev. 25, October 2017 XVI-30 NOZZLE SPRAY TEST - PRESSURE DROP OF 80 PSIG XVI-31 NOZZLE SPRAY TEST - PRESSURE DROP OF 80 PSIG XVI-32 NOZZLE SPRAY TEST - PRESSURE DROP OF 30 PSIG XVI-33 NOZZLE SPRAY TEST - PRESSURE DROP OF 30 PSIG XVI-34 SEISMIC ANALYSIS - REACTOR BUILDING XVI-35 DYNAMIC ANALYSIS - DRYWELL XVI-36 REACTOR SUPPORT STRUCTURE - SEISMIC XVI-37 SEISMIC ANALYSIS - WASTE BUILDING XVI-38 SEISMIC ANALYSIS - SCREENHOUSE XVI-39 SEISMIC ANALYSIS - TURBINE BUILDING (NORTH OF ROW C) XVI-40 SEISMIC ANALYSIS - TURBINE BUILDING (SOUTH OF ROW C) XVI-41 SEISMIC ANALYSIS - CONCRETE VENTILATION STACK XVI-42 REACTOR BUILDING MATHEMATICAL MODEL (NORTH-SOUTH) XVI-43 REACTOR SUPPORT STRUCTURE - SEISMIC XVI-44 REACTOR SUPPORT STRUCTURE - REACTOR BUILDING XVI-45 REACTOR SUPPORT STRUCTURE - REACTOR BUILDING AND SEISMIC XVI-46 PLAN OF BUILDING XVI-47 WALL SECTION 1 XVI-48 WALL SECTION 1 - DETAIL "A" XVI-49 WALL SECTION 1 - DETAIL "B" NMP Unit 1 UFSAR Figure Number Title TOC lxi Rev. 25, October 2017 XVI-50 WALL SECTION 1 - DETAIL "C" XVI-51 WALL SECTION 1 - DETAIL "D" XVI-52 WALL SECTION 1 - DETAIL "E" XVI-53 WALL SECTION 2 XVI-54 WALL SECTION 3 XVI-55 WALL SECTION 3A - DETAILS XVI-56 WALL SECTION 4 XVI-57 WALL SECTION 4 - DETAIL 1 XVI-58 WALL SECTION 4 - DETAIL 2 XVI-59 WALL SECTION 5 XVI-60 WALL SECTION 6 XVI-61 WALL SECTION 7 XVII-1 AVERAGE WIND ROSES FOR JANUARY '63-'64 XVII-2 AVERAGE WIND ROSES FOR FEBRUARY '63-'64 XVII-3 AVERAGE WIND ROSES FOR MARCH '63-'64 XVII-4 AVERAGE WIND ROSES FOR APRIL '63-'64 XVII-5 AVERAGE WIND ROSES FOR MAY '63-'64 XVII-6 AVERAGE WIND ROSES FOR JUNE '63-'64 XVII-7 AVERAGE WIND ROSES FOR JULY '63-'64 XVII-8 AVERAGE WIND ROSES FOR AUGUST '63-'64 XVII-9 AVERAGE WIND ROSES FOR SEPTEMBER '63-'64 NMP Unit 1 UFSAR Figure Number Title TOC lxii Rev. 25, October 2017 XVII-10 AVERAGE WIND ROSES FOR OCTOBER '63-'64 XVII-11 AVERAGE WIND ROSES FOR NOVEMBER '63-'64 XVII-12 AVERAGE WIND ROSES FOR DECEMBER '63-'64 XVII-13 AVERAGE WIND ROSES FOR '63-'64 XVII-14 AVERAGE DIURNAL LAPSE RATE JANUARY '63-'64, FEBRUARY '63-'64 XVII-15 AVERAGE DIURNAL LAPSE RATE MARCH '63-'64, APRIL '63-'64 XVII-16 AVERAGE DIURNAL LAPSE RATE MAY '63-'64, JUNE '63-'64 XVII-17 AVERAGE DIURNAL LAPSE RATE JULY '63-'64, AUGUST '63-'64 XVII-18 AVERAGE DIURNAL LAPSE RATE SEPTEMBER '63-'64, OCTOBER '63-'64 XVII-19 AVERAGE DIURNAL LAPSE RATE NOVEMBER '63-'64, DECEMBER '62-'63 XVII-20 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JANUARY '63-'64 XVII-21 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR FEBRUARY '63-'64 XVII-22 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR MARCH '63-'64 XVII-23 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR APRIL '63-'64 XVII-24 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR MAY '63-'64 XVII-25 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JUNE '63-'64 NMP Unit 1 UFSAR Figure Number Title TOC lxiii Rev. 25, October 2017 XVII-26 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JULY '63-'64 XVII-27 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR AUGUST '63-'64 XVII-28 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR SEPTEMBER '63-'64 XVII-29 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR OCTOBER '63-'64 XVII-30 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR NOVEMBER '63-'64 XVII-31 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR DECEMBER '63-'64 XVII-32 SECTOR MAP XVII-33 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS I XVII-34 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS II XVII-35 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS III XVII-36 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS IV XVII-37 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS II BECOMING CLASS IV AT 2 KM AND CLASS II AT 23 KM XVII-38 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 16 KM XVII-39 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 2 KM XVII-40 RADIAL CONCENTRATIONS - TURBULENCE CLASS I XVII-41 RADIAL CONCENTRATIONS - TURBULENCE CLASS II XVII-42 RADIAL CONCENTRATIONS - TURBULENCE CLASS III NMP Unit 1 UFSAR Figure Number Title TOC lxiv Rev. 25, October 2017 XVII-43 RADIAL CONCENTRATIONS - TURBULENCE CLASS IV XVII-44 RADIAL CONCENTRATIONS - TURBULENCE CLASS II BECOMING CLASS IV AT 2 KM AND CLASS II AT 23 KM XVII-45 RADIAL CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 16 KM XVII-46 RADIAL CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 2 KM XVII-47 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS I XVII-48 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS II XVII-49 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS III XVII-50 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS IV XVII-51 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS II BECOMING CLASS IV AT 2 KM AND CLASS II AT 23 KM XVII-52 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS IV BECOMING CLASS II AT 16 KM XVII-53 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS IV BECOMING CLASS II AT 2 KM XVII-54 ASSUMED CONCENTRATION AND DOSE RATE DISTRIBUTIONS CLOSE TO THE ELEVATED SOURCE XVII-55 GAMMA DOSE RATE AS A FUNCTION OF y AT 1 KM FROM THE SOURCE XVII-56 SOUTHEASTERN LAKE ONTARIO XVII-57 DILUTION OF RISING PLUME XVII-58 ESTIMATED LAKE CURRENTS AT COOLING WATER DISCHARGE NMP Unit 1 UFSAR Figure Number Title TOC lxv Rev. 25, October 2017 XVII-59 TEMPERATURE PROFILES IN AN EASTWARD CURRENT AT THE OSWEGO CITY WATER INTAKE XVII-60 SUBSURFACE SECTION PLOT PLAN XVII-61 LOG OF BORING (BORING CB-1) XVII-62 LOG OF BORING (BORING CB-2) XVII-63 LOG OF BORING (BORING CB-3) XVII-64 LOG OF BORING (BORING CB-4) XVII-65 ATTENUATION CURVES

NMP Unit 1 UFSAR Section I EF I-1 Rev. 25, October 2017 LIST OF EFFECTIVE FIGURES SECTION I Figure Number Revision Number I-1 20 NMP Unit 1 UFSAR Section I i Rev. 25, October 2017 TABLE OF CONTENTS Section Title SECTION I INTRODUCTION AND SUMMARY A. PRINCIPAL DESIGN CRITERIA 1.0 General 2.0 Buildings and Structures 3.0 Reactor 4.0 Reactor Vessel 5.0 Containment 6.0 Control and Instrumentation 7.0 Electrical Power 8.0 Radioactive Waste Disposal 9.0 Shielding and Access Control 10.0 Fuel Handling and Storage B. CHARACTERISTICS 1.0 Site 2.0 Reactor 3.0 Core 4.0 Fuel Assembly 5.0 Control System 6.0 Core Design and Operating Conditions 7.0 Design Power Peaking Factor 8.0 Nuclear Design Data 9.0 Reactor Vessel 10.0 Coolant Recirculation Loops 11.0 Primary Containment 12.0 Secondary Containment 13.0 Structural Design 14.0 Station Electrical System 15.0 Reactor Instrumentation System 16.0 Reactor Protection System C. IDENTIFICATION OF CONTRACTORS D. GENERAL CONCLUSIONS E. REFERENCES NMP Unit 1 UFSAR Section I ii Rev. 25, October 2017 LIST OF TABLES Table Number Title I-1 COMPARISON TO STANDARDS - HISTORICAL (PROVIDED WITH APPLICATION TO CONVERT TO FULL-TERM OPERATING LICENSE) I-2 ABBREVIATIONS AND ACRONYMS USED IN UFSAR NMP Unit 1 UFSAR Section I iii Rev. 25, October 2017 LIST OF FIGURES Figure Number Title I-1 PIPING, INSTRUMENT AND EQUIPMENT SYMBOLS NMP Unit 1 UFSAR Section I I-1 Rev. 25, October 2017 SECTION I INTRODUCTION AND SUMMARY This report is submitted in accordance with 10 CFR Part 50.71(e) entitled "Periodic Updating of Final Safety Analysis Reports" for Nine Mile Point Nuclear Station - Unit 1 (Unit 1). The Station is located on the southeast shore of Lake Ontario, in Oswego County, New York, 7 mi northeast of the city of Oswego. The operating license (OL) was transferred to Nine Mile Point Nuclear Station, LLC (NMPNS), on November 7, 2001, under License Amendment No. 172. A. PRINCIPAL DESIGN CRITERIA The following paragraphs describing the principal design criteria are oriented toward the twenty-seven criteria issued by the United States Atomic Energy Commission (USAEC) on November 22, 1965.(1) The twenty-seven criteria represented proposed "General Design Criteria for Nuclear Power Plant Construction Permits." The twenty-seven criteria are presented here for historical reference and are followed by the Unit 1 principal design criteria. Table I-1 provides historical information regarding an assessment of Unit 1 against criteria that were being used by the USAEC at the time of the Unit 1 application for a full-term OL. Facility Criterion 1 Those features of reactor facilities which are essential to the prevention of accidents or to the mitigation of their consequences must be designed, fabricated, and erected to: (a) Quality standards that reflect the importance of the safety function to be performed. It should be recognized, in this respect, that design codes commonly used for nonnuclear applications may not be adequate. NMP Unit 1 UFSAR Section I I-2 Rev. 25, October 2017 (b) Performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces imposed by the most severe earthquakes, flooding conditions, winds, ice, and other natural phenomena anticipated at the proposed site. Criterion 2 Provisions must be included to limit the extent and the consequences of credible chemical reactions that could cause or materially augment the release of significant amounts of fission products from the facility. Criterion 3 Protection must be provided against possibilities for damage of the safeguarding features of the facility by missiles generated through equipment failures inside the containment. Reactor Criterion 4 The reactor must be designed to accommodate, without fuel failure or primary system damage, deviations from steady state norm that might be occasioned by abnormal yet anticipated transient events such as tripping of the turbine generator and loss of power to the reactor recirculation system pumps. Criterion 5 The reactor must be designed so that power or process variable oscillations or transients that could cause fuel failure or primary system damage are not possible or can be readily suppressed. Criterion 6 Clad fuel must be designed to accommodate throughout its design lifetime all normal and abnormal modes of anticipated reactor operation, including the design overpower condition, without experiencing significant cladding failures. Unclad or vented fuels must be designed with the similar objective of providing control over NMP Unit 1 UFSAR Section I I-3 Rev. 25, October 2017 fission products. For unclad and vented solid fuels, normal and abnormal modes of anticipated reactor operation must be achieved without exceeding design release rates of fission products from the fuel over core lifetime. Criterion 7 The maximum reactivity worth of control rods or elements and the rates with which reactivity can be inserted must be held to values such that no single credible mechanical or electrical control system malfunction could cause a reactivity transient capable of damaging the primary system or causing significant fuel failure. Criterion 8 Reactivity shutdown capability must be provided to make and hold the core subcritical from any credible operating condition with any one control element at its position of highest reactivity. Criterion 9 Backup reactivity shutdown capability must be provided that is independent of normal reactivity control provisions. This system must have the capability to shut down the reactor from any operating condition. Criterion 10 Heat removal systems must be provided which are capable of accommodating core decay heat under all anticipated abnormal and credible accident conditions, such as isolation from the main condenser and complete or partial loss of primary coolant from the reactor. Criterion 11 Components of the primary coolant and containment systems must be designed and operated so that no substantial pressure or thermal stress will be imposed on the structural materials unless the temperatures are well above the nil ductility temperatures. For ferritic materials of the coolant envelope and the containment, minimum temperatures are NDT + 60°F and NDT + 30°F, respectively. Criterion 12 NMP Unit 1 UFSAR Section I I-4 Rev. 25, October 2017 Capability for control rod insertion under abnormal conditions must be provided. Criterion 13 The reactor facility must be provided with a control room from which all actions can be controlled or monitored as necessary to maintain safe operational status of the plant at all times. The control room must be provided with adequate protection to permit occupancy under the conditions described in Criterion 17 below, and with the means to shut down the plant and maintain it in a safe condition if such accident were to be experienced. Criterion 14 Means must be included in the control room to show the relative reactivity status of the reactor such as position indication of mechanical rods or concentrations of chemical poisons. Criterion 15 A reliable reactor protection system must be provided to automatically initiate appropriate action to prevent safety limits from being exceeded. Capability must be provided for testing functional operability of the system and for determining that no component or circuit failure has occurred. For instruments and control systems in vital areas where the potential consequences of failure require redundancy, the redundant channels must be independent and must be capable of being tested to determine that they remain independent. Sufficient redundancy must be provided that failure or removal from service of a single component or channel will not inhibit necessary safety action when required. These criteria should, where applicable, be satisfied by the instrumentation associated with containment closure and isolation systems, afterheat removal and core cooling systems, systems to prevent cold-slug accidents, and other vital systems, as well as the reactor nuclear and process safety system. Criterion 16 The vital instrumentation systems of Criterion 15 must be designed so that no credible combination of circumstances NMP Unit 1 UFSAR Section I I-5 Rev. 25, October 2017 can interfere with the performance of a safety function when it is needed. In particular, the effect of influences common to redundant channels which are intended to be independent must not negate the operability of a safety system. The effects of gross disconnection of the system, loss of energy (electric power, instrument air), and adverse environment (heat from loss of instrument cooling, extreme cold, fire, steam, water, etc.) must cause the system to go into its safest state (fail-safe) or be demonstrably tolerable on some other basis. Engineered Safeguards Criterion 17 The containment structure, including access openings and penetrations, must be designed and fabricated to accommodate or dissipate without failure the pressures and temperatures associated with the largest credible energy release including the effects of credible metal-water or other chemical reactions uninhibited by active quenching systems. If part of the primary coolant system is outside the primary reactor containment, appropriate safeguards must be provided for that part if necessary, to protect the health and safety of the public, in case of an accidental rupture in that part of the system. The appropriateness of safeguards such as isolation valves, additional containment, etc., will depend on environmental and population conditions surrounding the site. Criterion 18 Provisions must be made for the removal of heat from within the containment structure as necessary to maintain the integrity of the structure under the conditions described in Criterion 17 above. If engineered safeguards are needed to prevent containment vessel failure due to heat released under such conditions, at least two independent systems must be provided, preferably of different principles. Backup equipment (e.g., water and power systems) to such engineered safeguards must also be redundant. Criterion 19 The maximum integrated leakage from the containment structure under the conditions described in Criterion 17 above must meet the site exposure criteria set forth in NMP Unit 1 UFSAR Section I I-6 Rev. 25, October 2017 10CFR100. The containment structure must be designed so that the containment can be leak tested at least to design pressure conditions after completion and installation of all penetrations, and the leakage rate measured over a suitable period to verify its conformance with required performance. The plant must be designed for later tests at suitable pressures. Criterion 20 All containment structure penetrations subject to failure such as resilient seals and expansion bellows must be designed and constructed so that leak-tightness can be demonstrated at design pressure at any time throughout operating life of the reactor. Criterion 21 Sufficient normal and emergency sources of electrical power must be provided to assure a capability for prompt shutdown and continued maintenance of the reactor facility in a safe condition under all credible circumstances. Criterion 22 Valves and their associated apparatus that are essential to the containment function must be redundant and so arranged that no credible combination of circumstances can interfere with their necessary functioning. Such redundant valves and associated apparatus must be independent of each other. Capability must be provided for testing functional operability of these valves and associated equipment to determine that no failure has occurred and that leakage is within acceptable limits. Redundant valves and auxiliaries must be independent. Containment closure valves must be actuated by instrumentation, control circuits and energy sources which satisfy Criterion 15 and 16 above. Criterion 23 In determining the suitability of a facility for a proposed site the acceptance of the inherent and engineered safety afforded by the systems, materials and components, and the associated engineered safeguards built into the facility, will depend on their demonstrated performance capability and reliability and the extent to which the operability of such systems, materials, components, and engineered NMP Unit 1 UFSAR Section I I-7 Rev. 25, October 2017 safeguards can be tested and inspected during the life of the plant. Radioactivity Control Criterion 24 All fuel storage and waste handling systems must be contained if necessary to prevent the accidental release of radioactivity in amounts which could affect the health and safety of the public. Criterion 25 The fuel handling and storage facilities must be designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel under all anticipated normal and abnormal conditions, and credible accident conditions. Variables upon which health and safety of the public depend must be monitored. Criterion 26 Where unfavorable environmental conditions can be expected to require limitations upon the release of operational radioactive effluents to the environment, appropriate holdup capacity must be provided for retention of gaseous, liquid, or solid effluents. Criterion 27 The plant must be provided with systems capable of monitoring the release of radioactivity under accident conditions. 1.0 General The Station is intended as a high load factor generating facility. The recirculation flow control system described in Section VIII contributes to this objective by providing a relatively fast means for adjusting the Station output over a preselected power range. Overall reliability, routine and periodic test requirements, and other design considerations must also be compatible with this objective. Careful attention has been given to fabrication procedures and adherence to Code requirements. The rigid requirements of NMP Unit 1 UFSAR Section I I-8 Rev. 25, October 2017 specific portions of various codes have been arbitrarily applied to some safety-related systems to ensure quality construction in such cases where the complete Code does not apply. For piping, the ASA B31.1-1955 Code was used and where exceptions were taken, safety evaluations were performed to document that an adequate margin of safety was maintained. Periodic test programs have been developed for required engineered safeguards equipment. These tests cover component testing such as pumps and valves and full system tests, duplicating as closely as possible the accident conditions under which a given system must perform. 2.0 Buildings and Structures The Station plot plan, design and arrangement of the various buildings and structures are described in Section III. Principal structures and equipment which may serve either to prevent accidents or to mitigate their consequences are designed, fabricated and erected in accordance with applicable codes to withstand the most severe earthquake, flooding condition, windstorm, ice condition, temperature and other deleterious natural phenomena which can be expected to occur at the site. 3.0 Reactor 1. A direct-cycle boiling water system reactor (BWR), described in Section IV, is employed to produce steam (1030 psig in reactor vessel, 950 psig turbine inlet) for use in a steam-driven turbine generator. The rated thermal output of the reactor is 1850 MWt. 2. The reactor is fueled with slightly enriched uranium dioxide contained in Zircaloy clad fuel rods described in Section IV. Selected fuel rods also incorporate small amounts of gadolinium as burnable poison. 3. To avoid fuel damage, the minimum critical power ratio (MCPR) is maintained greater than or equal to the safety limit CPR. 4. The fuel rod cladding is designed to maintain its integrity throughout the anticipated fuel life as described in Section IV. Fission gas release within NMP Unit 1 UFSAR Section I I-9 Rev. 25, October 2017 the rods and other factors affecting design life are considered for the maximum expected burnup. 5. The reactor and associated systems are designed so that there is no inherent tendency for undamped oscillations. A stability analysis evaluation is given in Section IV. 6. Heat removal systems are provided which are capable of safely accommodating core decay heat under all credible circumstances, including isolation from the main condenser and loss of coolant from the reactor. Each different system so provided has appropriate redundant features. Independent auxiliary cooling means are provided to cool the reactor under a variety of conditions. The normal auxiliary cooling means during shutdown and refueling is the shutdown cooling system described in Section X-A. A redundant emergency cooling system, described in Section V-E, is provided to remove decay heat in the event the reactor is isolated from the main condenser while still under pressure. Additional cooling capability is also available from the high-pressure coolant injection (HPCI) system and the fire protection system. Redundant and independent core spray systems are provided to cool the core in the event of a loss-of-coolant accident (LOCA). Automatic depressurization is included to rapidly reduce pressure to assist with core spray operation (see Section VII-A). Operation of the core spray system assures that any metal-water reaction following a postulated LOCA will be limited to less than 1 percent of the Zircaloy clad. 7. Reactivity shutdown capability is provided to make and hold the core adequately subcritical, by control rod action, from any point in the operating cycle and at any temperature down to room temperature, assuming that any one control rod is fully withdrawn and unavailable for use. This capability is demonstrated in Section IV-B. A physical description of the movable control rods is NMP Unit 1 UFSAR Section I I-10 Rev. 25, October 2017 given in Section IV-B. The control rod drive (CRD) hydraulic system is described in Section X-C. The force available to scram a control rod is approximately 3000 lb at the beginning of a scram stroke. This is well in excess of the 440-lb force required in the event of fuel channel pinching of the control rod blade during a LOCA, as discussed in Section XV. Even with scram accumulator failure, a force of at least 1100 lb from reactor pressure acting alone is available with reactor pressures in excess of 800 psig. 8. Redundant reactivity shutdown capability is provided independent of normal reactivity control provisions. This system has the capability, as shown in Section VII-C, to bring the reactor to a cold shutdown condition at any time in the core life, independent of the control rod system capabilities. Cycle-specific results are contained in the SRLR(2). 9. A flow restrictor in the main steam line (MSL) limits coolant loss from the reactor vessel in the event of a MSL break (Section VII-F). 4.0 Reactor Vessel 1. The reactor core and vessel are designed to accommodate tripping of the turbine generator, loss of power to the reactor recirculation system and other transients, and maneuvers which can be expected without compromising safety and without fuel damage. A bypass system having a design capacity of approximately 40 percent of turbine steam flow for the throttle valves wide open (VWO) condition partially mitigates the effects of sudden load rejection. An actual bypass system test was performed and the results indicated a system bypass capacity of about 2,500,000 lb/hr. This and other transients and maneuvers which have been analyzed are detailed in Section XV. 2. Separate systems to prevent serious reactor coolant system (RCS) overpressure are incorporated in the design. These include an overpressure scram, solenoid-actuated relief valves, safety valves and the NMP Unit 1 UFSAR Section I I-11 Rev. 25, October 2017 turbine bypass system. An analysis of the adequacy of RCS pressure relief devices is included in Section V-C. 3. Power excursions which could result from any credible reactivity addition accident will not cause damage, either by motion or rupture, to the pressure vessel, or impair operation of required safeguards systems. The magnitude of credible reactivity addition accidents is curtailed by control rod velocity limiters (Section VII-D), by a control rod housing support structure (Section VII-E), and by procedural controls supplemented by a rod worth minimizer (RWM) (Section VIII-C). Power excursion analyses for control rod dropout accidents are included in Section XV. 4. The reactor vessel will not be substantially pressurized until the vessel wall temperature is in excess of the nil ductility reference temperature (RTNDT) + 60°F. The initial RTNDT of the reactor vessel material is no greater than 40°F. The change of RTNDT with radiation exposure has been evaluated in accordance with Regulatory Guide (RG) 1.99 Revision 2 to determine an adjusted reference temperature (ART) for the most limiting vessel material. Vessel material surveillance samples are located within the reactor vessel to permit periodic verification of material properties with exposure. 5.0 Containment 1. The primary containment, including the drywell, pressure suppression chamber, and associated access openings and penetrations, is designed, fabricated and erected to accommodate, without failure, the pressures and temperatures resulting from or subsequent to the double-ended rupture (DER) or equivalent failure of any coolant pipe within the drywell. The primary containment is designed to accommodate the pressures following a LOCA, including the generation of hydrogen from a metal-water reaction. Pressure transients, including hydrogen effects, are presented in Section XV. NMP Unit 1 UFSAR Section I I-12 Rev. 25, October 2017 The initial NDTT for the primary containment system is about -20°F and is not expected to increase during the lifetime of the Station. These structures are described in Sections VI-A, B and C. Additional details, particularly those related to design and fabrication, are included in Section XVI. 2. Provisions are made for the removal of heat from within the primary containment, for reasonable protection of the containment from fluid jets or missiles, and such other measures as may be necessary to maintain the integrity of the containment system as long as necessary following a LOCA. Redundant containment spray systems, described in Section VII, pump water from the suppression chamber through independent heat exchangers to spray nozzles, which discharge into the drywell and suppression chamber. Water sprayed into the drywell is returned by gravity to the suppression chamber to complete the cooling cycle. Studies performed to verify the capability of the containment system to withstand potential fluid jets and missiles are summarized in Section XVI. 3. Provision is made for periodic integrated leakage rate tests (ILRT) to be performed in accordance with 10CFR50 Appendix J. Provision is also made for leak testing penetrations and access openings and for periodically demonstrating the integrity of the reactor building. These provisions are all described in Section VI-F. 4. The containment system and all other necessary engineered safeguards were originally designed and maintained such that offsite doses resulting from postulated accidents are below the values stated in 10CFR100. The offsite doses have been re-analyzed in accordance with RG 1.183 and 10CFR50.67. The analysis results are detailed in Section XV. 5. Double isolation valves are provided on most lines directly entering the primary containment freespace, or penetrating the primary containment and connected to the RCS. Lines which are not equipped with double isolation valves have been determined to be acceptable NMP Unit 1 UFSAR Section I I-13 Rev. 25, October 2017 based upon the fact that the system reliability is not compromised, the system is closed outside containment, and a single active failure can be accommodated with only one isolation valve in the line. Periodic testing of these valves will assure their capability to isolate at all times. The isolation valve system is discussed in detail in Section VI-D. 6.The reactor building provides secondary containment when the pressure suppression system is in service and serves as the primary containment barrier during refueling and other periods when the pressure suppression system is open or not required. This structure is described in Section VI-C. An emergency ventilation system (Section VII-H) provides a means for controlled release of halogens and particulates via filters from the reactor building to the stack under accident conditions. 6.0 Control and Instrumentation 1. The Station is provided with a control room (Section III-B) which has adequate shielding and other emergency features to permit occupancy during all credible accident situations. 2. Interlocks or other protective features are provided to augment the reliability of procedural controls in preventing serious accidents. Interlock systems are provided which block or prevent rod withdrawal from a multitude of abnormal conditions. The control rod block logic is shown on Figures VIII-6 and VIII-8, respectively, for the source range monitor (SRM) and intermediate range monitor (IRM) neutron instrumentation. In the power range, average power range monitor (APRM) instrumentation provides both control rod and recirculation flow control blocks, as shown on Figure VIII-14. Reactivity excursions involving the control rods are either prevented or their consequences substantially mitigated by a control RWM (Section VIII-C.4.0) which supplements procedural controls in avoiding patterns of high rod worths, a local power range monitor (LPRM) neutron monitoring and alarm system (Section NMP Unit 1 UFSAR Section I I-14 Rev. 25, October 2017 VIII-C.1.1.3), and a control rod position indicating system (Section IV-B.6.0), both of which enable the Operator to observe rod movement, thus verifying his actions. A control rod overtravel position light verifies that the blade is coupled to a withdrawn CRD. A refueling platform operation interlock is discussed in Section XV, Refueling Accident, which, along with other procedures and supplemented by automatic interlocks, serves to prevent criticality accidents in the refueling mode. A cold water addition reactivity excursion is prevented by the procedures and interlocks described in Section XV, Startup of Cold Recirculation Loop (Transient Analysis). Containment integrity is maintained through the use of strict procedural controls and is enforced by interlocking mechanisms at the airlock doors to the drywell and a local alarm system at the access openings of the reactor building. 3. A reliable, dual-logic channel reactor protection system (RPS), described in Section VIII-A, is provided to automatically initiate appropriate action whenever various parameters exceed preset limits. Each logic channel contains two subchannels with completely independent sensors, each capable of tripping the logic channel. A trip of one-of-two subchannels in each logic channel results in a reactor scram. The trip in each logic channel may occur from unrelated parameters, i.e., high neutron flux in one logic channel coupled with high pressure in the other logic channel will result in a scram. The RPS circuitry fails in a direction to cause a reactor scram in the event of loss of power or loss of air supply to the scram solenoid valves. Periodic testing and calibration of individual subchannels is performed to assure system reliability. The ability of the RPS to safely terminate a variety of Station malfunctions is demonstrated in Section XV. 4. Redundant sensors and circuitry are provided for the actuation of equipment required to function under post-accident conditions. This redundancy is NMP Unit 1 UFSAR Section I I-15 Rev. 25, October 2017 described in the various sections of the text discussing system design. 7.0 Electrical Power Sufficient normal and standby auxiliary sources of electrical power are provided to assure a capability for prompt shutdown and continued maintenance of the Station in a safe condition under all credible circumstances. These features are discussed in Section IX. 8.0 Radioactive Waste Disposal 1. Gaseous, liquid and solid waste disposal facilities are designed so that discharge of effluents is in accordance with 10CFR20 and 10CFR50 Appendix I. The facility descriptions are given in Section XII-A while the development of appropriate limits is covered in Section II. 2. Gaseous discharge from the Station is appropriately monitored, as discussed in Section VIII, and automatic isolation features are incorporated to maintain releases below the limits of 10CFR20 and 10CFR50 Appendix I. 9.0 Shielding and Access Control Radiation shielding and access control patterns are such that doses will be less than those specified in 10CFR20. These features are described in Section XII-B. 10.0 Fuel Handling and Storage Appropriate fuel handling and storage facilities which preclude accidental criticality and provide adequate cooling for spent fuel are described in Section X. B. CHARACTERISTICS The following is a summary of design and operating characteristics. 1.0 Site Location Oswego County, New York State Size of Site 900 Acres NMP Unit 1 UFSAR Section I I-16 Rev. 25, October 2017 Site and Station Nine Mile Point Nuclear Station, Ownership LLC (NMPNS) Net Electrical Output 615 MW (Maximum) 2.0 Reactor Reference Rated Thermal 1850 MW Output Dome Pressure 1030 psig Turbine Inlet Pressure 950 psig Total Core Coolant 67.5 x 106 lb/hr Flow Rate Steam Flow Rate 7.32 x 106 lb/hr 3.0 Core Circumscribed Core 167.16 in Diameter Active Core 171.125 in Height + Assembly 4.0 Fuel Assembly Number of Fuel Assemblies 532 Fuel Rod Array SRLR(2) Fuel Rod Pitch References 3 and 4 Cladding Material References 3 and 4 Fuel Material UO2 and UO2-Gd2O3 Active Fuel Length References 3 and 4 Cladding Outside Diameter References 3 and 4 Cladding Thickness References 3 and 4 Fuel Channel Material References 3 and 4 5.0 Control System Number of Movable Control 129 Rods Shape of Movable Control Cruciform Rods Pitch of Movable Control 12.0 in Rods Control Material in B4C - 70% Theoretical Movable Control Rods Density; Hafnium Type of Control Drives Bottom Entry, Hydraulic Actuated Control of Reactor Output Movement of Control Rods and Variation of Coolant Flow Rate NMP Unit 1 UFSAR Section I I-17 Rev. 25, October 2017 6.0 Core Design and Operating Conditions Maximum Linear Heat Core Operating Limits Report Generation Rate Heat Transfer Surface

• Area Average Heat Flux - Rated
• Power Minimum Critical Power Core Operating Limits Report Ratio for Most Limiting Transients Core Average Void Fraction - Coolant within Assemblies Core Average Exit
• Quality - Coolant within Assemblies 7.0 Design Power Peaking Factor Total Peaking Factor Core Operating Limits Report 8.0 Nuclear Design Data Average Initial Volume References 3 and 5 Metric Enrichment Beginning of Cycle - Core Effective Multiplication and Control System Worth - No Voids, 20C(2) Uncontrolled SRLR(2) Fully Controlled SRLR(2) Strongest Control SRLR(2) Rod Out Standby Liquid Control System Capability: These parameters are recalculated for each reload because of their dependency on core composition and exposure. These calculated values are intermediate quantities that do not represent design requirements or operating limits and thus are not separately reported in the SRLR(2). These parameters are recalculated for each reload because of their dependency on core composition and exposure. These calculated values are intermediate quantities that do not represent design requirements or operating limits and thus are not separately reported in the SRLR(2). Total peaking is monitored with LHGR limits.

NMP Unit 1 UFSAR Section I I-18 Rev. 25, October 2017 Shutdown Margin (k) ppm (20C, Xenon Free) SRLR(2) SRLR(2) 9.0 Reactor Vessel Inside Diameter 17 ft - 9 in Internal Height 63 ft - 10 in Design Pressure 1250 psig at 575°F 10.0 Coolant Recirculation Loops Location of Recirculation Containment Drywe1l Loops Number of Recirculation 5 Loops and Pumps Pipe Size 28 in 11.0 Primary Containment Type Pressure Suppression Design Pressure of 62 psig Drywell Vessel Design Pressure of 35 psig Suppression Chamber Vessel Design Leakage Rate 0.5 weight percent per day at 35 psig 12.0 Secondary Containment Type Reinforced concrete and steel superstructure with metal siding Internal Design Pressure 40 lb/ft2 Design Leakage Rate 100% free volume per day discharged via stack while maintaining 0.25-in water negative pressure in the reactor building relative to atmosphere 13.0 Structural Design Seismic Ground 0.11g Acceleration Sustained Wind Loading 125 mph, 30 ft above ground level NMP Unit 1 UFSAR Section I I-19 Rev. 25, October 2017 Control Room Shielding Normal Operation - Dose not to exceed hourly equivalent (based on 40-hr week) of maximum permissible quarterly dose specified in 10CFR20. Accident Conditions - Meets the design total effective dose equivalent (TEDE) dose for personnel in the control room such that the exposure limits of 10CFR50.67 will not be exceeded in the course of the LOCA. In addition, the cumulative dose from any design basis accident (DBA) would also meet 10CFR50.67 limits. 14.0 Station Electrical System Incoming Power Sources Two 115-kV transmission lines Outgoing Power Lines Two 345-kV transmission lines Onsite Power Sources Two diesel generators Provided Two safety-related Station batteries One Q-related 125-V dc battery system 15.0 Reactor Instrumentation System Location of Neutron In-core Monitor Sensors Ranges of Nuclear Instrumentation: Four Startup Range Source to 0.01% rated power and to Monitors 8.3% with chamber retraction Eight Intermediate Range 0.0003% to 40% rated power Monitors 120 Power Range Monitors 5% to 125% rated power 16.0 Reactor Protection System Number of Channels in 2 Reactor Protection System Number of Channels 2 NMP Unit 1 UFSAR Section I I-20 Rev. 25, October 2017 Required to Scram or Effect Other Protective Functions Number of Sensors per 2 Monitored Variable in each Channel (Minimum for scram function) C. IDENTIFICATION OF CONTRACTORS The General Electric Company (GE) was engaged to design, fabricate and deliver the nuclear steam supply system (NSSS), turbine generator, and other major elements and systems. GE also furnished the complete core design and nuclear fuel supply for the initial core. Global Nuclear Fuel (GNF) is currently furnishing replacement cores. Niagara Mohawk Power Corporation (NMPC), acting as the architect-engineer, specified and procured the remaining systems and components, including the pressure suppression containment system, and coordinated the complete integrated Station. Stone and Webster Engineering Corporation (SWEC) was engaged to manage field construction. Currently, various contractors are utilized to assist in continuous Station modifications. D. GENERAL CONCLUSIONS The favorable site characteristics, criteria and design requirements of all the systems related to safety, the potential consequences of postulated accidents, and the technical competence of the applicant and its contractors, assure that Unit 1 can be operated without endangering the health and safety of the public. E. REFERENCES 1. USAEC Press Release H-252, "General Design Criteria for Nuclear Power Plant Construction Permits," November 22, 1965. 2. "Supplimental Reload Licensing Report for Nine Mile Point 1 Reload 24 Cycle 25," 002N6949, Revision 0, March 2017. 3. "Global Nuclear Fuels Fuel Bundle Designs," NEDE-31152P, Revision 9, May 2007. NMP Unit 1 UFSAR Section I I-21 Rev. 25, October 2017 4. "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II)," NEDC-33270P, Revision 7, October 2016. 5. "Fuel Bundle Information Report for Nine Mile Point 1 Reload 24 Cycle 25," 002N6950, Revision 0, March 2017. NMP Unit 1 UFSAR Section I I-22 Rev. 25, October 2017 TABLE I-1 COMPARISON TO STANDARDS-HISTORICAL (PROVIDED WITH APPLICATION TO CONVERT TO FULL-TERM OPERATING LICENSE) As part of its application to convert to a full-term operating license, NMPC provided an assessment of Unit 1 against criteria being used by the Commission in evaluating new plants. Construction of Unit 1 was well along or already completed when many of these standards were developed. These assessments discussed the adequacy of Unit 1 in relation to Appendices A through J of 10CFR50, Safety Guides 1 through 21, IEEE Standards, and Regulatory Guides 1.22 through 1.59. This historical information is located in the permanent plant file under the following submittals: Technical Supplement to Petition for Conversion From Provisional Operating License to Full-Term Operating License, July 1972 Amendment No. 1 to Application to Convert Provisional Operating License to Full-Term Operating License, November 1973 The information provided in the above submittals does not represent a commitment to the standards. NMP Unit 1 UFSAR Section I I-23 Rev. 25, October 2017 TABLE I-2 ABBREVIATIONS AND ACRONYMS USED IN UFSAR ACI American Concrete Institute ADS Automatic depressurization system AISC American Institute of Steel Construction ALARA As low as reasonably achievable ALRA Amended license renewal application AMP Aging Management Program ANS American Nuclear Society ANSI American National Standards Institute AOO Anticipated operational occurrence AOV Air-operated valve APRM Average power range monitor ARI Alternate rod injection ARMS Area radiation monitoring system ART Adjusted reference temperature AST Alternative source term ASTM American Society for Testing and Materials ATWS Anticipated transient without scram BOC Beginning of cycle BOP Balance of plant BPWS Banked position withdrawal sequence BTP Branch technical position BWR Boiling water reactor BWROG Boiling Water Reactor Owners' Group BWRT Backwash receiving tank BWRVIP Boiling Water Reactor Vessel and Internals Program CAD Containment atmosphere dilution (device) CCCWS Closed-cycle cooling water system CEO Chief Executive Officer CFR Code of Federal Regulations CFS Condensate filtration system CGCS Combustible gas control system CHF Critical heat flux CIV Combined intermediate valve CND Condensate demineralizer CO2 Carbon dioxide COLR Core Operating Limits Report CPR Critical power ratio CRD Control rod drive CRDA Control rod drop accident CRDRL Control rod drive return line NMP Unit 1 UFSAR Section I I-24 Rev. 25, October 2017 TABLE I-2 (Cont'd.) CRPI Control rod position indication CRS Control Room Supervisor CRT Cathode ray tube CSB Cold Storage Building CSO Chief Shift Operator CST Condensate storage tank CUF Cumulative usage factor CWT Concentrated waste tank DAC Dominant area of concern DBA Design basis accident DBE Design basis earthquake DCRDR Detailed control room design review DE Dose equivalent DEC Department of Environmental Conservation DER Deviation/Event Report DER Double-ended rupture DG Diesel generator DOP Dioctylphthalate DOT Department of Transportation ECCS Emergency core cooling system ECP Electrochemical corrosion potential EDG Emergency diesel generator EFPY Effective full-power years EIC Energy Information Center EOC End of cycle EOF Emergency Operations Facility EOL End of life EOP Emergency operating procedure EPA Environmental Protection Agency EPDM Ethylene-propylene-diene-monomer EPG Emergency procedure guideline EPRI Electric Power Research Institute EQ Environmental qualification ESF Engineered safety feature ESW Emergency service water FA Fire area FAC Flow-accelerated corrosion FCV Flow control valve FHA Fire Hazards Analysis FMEA Failure modes and effects analysis FMP Fatigue Monitoring Program FRC Franklin Research Center NMP Unit 1 UFSAR Section I I-25 Rev. 25, October 2017 TABLE I-2 (Cont'd.) FSA Fire subarea FSAR Final Safety Analysis Report FZ Fire zone GALL Generic aging lessons learned GDC General Design Criterion GE General Electric Company GL Generic Letter GSI Generic Safety Issue HAZ Heat-affected zone HCU Hydraulic control unit HEM Homogeneous equilibrium model HEO Human engineering observation HEPA High-efficiency particulate air/absolute (filter) HPCI High-pressure coolant injection HVAC Heating, ventilating, and air conditioning HWC Hydrogen water chemistry HX Heat exchanger I&C Instrumentation & control ID Inner diameter IGSCC Intergranular stress corrosion cracking ILRT Integrated leakage rate test INPO Institute of Nuclear Power Operations ISB ISFSI Storage Building ISEG Independent Safety Engineering Group ISFSI Independent Spent Fuel Storage Installation ISI Inservice inspection ISP Integrated Surveillance Program IST Inservice testing LCO Limiting condition of operation LHGR Linear heat generation rate LLD Lower limit of detection LLL Low-low limit LOCA Loss-of-coolant accident LOFW Loss of feedwater LOOP Loss of offsite power LPCS Low-pressure core spray LPRM Local power range monitor LPSP Low power setpoint LPZ Low population zone LRA License renewal application LSSS Limiting safety system setting NMP Unit 1 UFSAR Section I I-26 Rev. 25, October 2017 TABLE I-2 (Cont'd.) LTC Load tap changer M&TE Measuring and testing equipment MAPLHGR Maximum average planar linear heat generation rate MCC Motor control center MCPR Minimum critical power ratio MG Motor generator MLHGR Maximum linear heat generation rate MOV Motor-operated valve MSIV Main steam isolation valve MSL Main steam line MSLB Main steam line break NDT Nil ductility transition NDT Nondestructive testing NDTT Nil ductility transition temperature NEI Nuclear Energy Institute NEIL Nuclear Electric Insurance Limited NFI New fuel introduction NFPA National Fire Protection Association NMPC Niagara Mohawk Power Corporation NMPNS Nine Mile Point Nuclear Station NPSH Net positive suction head NRC Nuclear Regulatory Commission NRV Nonreturn valve NSSS Nuclear steam supply system NVLAP National Voluntary Laboratory Accreditation Program NYPA New York Power Authority NYPP New York Power Pool OBE Operating basis earthquake OCCWS Open-cycle cooling water system OEA Operating experience assessment OL Operating license OLNC On-Line NobleChem OOS Out of service OSC Operational Support Center OT Operational transient PA Public address (system) PASS Post-accident sampling system PCI Pellet-cladding interaction NMP Unit 1 UFSAR Section I I-27 Rev. 25, October 2017 TABLE I-2 (Cont'd.) PCT Peak cladding temperature p.f. Power factor P&ID Piping and instrumentation diagram PIV Pressure isolation valve PM Preventive maintenance PORC Plant Operations Review Committee PP/PA Page party/public address (system) PSAR Preliminary Safety Analysis Report PSTG Plant-specific technical guideline P-T Pressure-temperature PTLR Pressure and Temperature Limits Report PVC Polyvinyl chloride QA Quality assurance QATR Quality Assurance Topical Report RBCLCW Reactor building closed loop cooling water RBM Rod block monitor RCA Radiologically-controlled area RCPB Reactor coolant pressure boundary RCS Reactor coolant system RG Regulatory Guide RIP Reactor internals protection RIS Regulatory Issue Summary RLO Reduced loop operation RMS Radiation monitoring system RO Reactor Operator RPIS Rod position information system RPS Reactor protection (trip) system RPT Recirculation pump trip RPV Reactor pressure vessel RPVH Reactor pressure vessel head RSP Remote shutdown panel RSS Remote shutdown system RTD Resistance temperature detector RTNDT Reference temperature nil ductility transition RWCU Reactor water cleanup RWE Rod withdrawal error RWM Rod worth minimizer RWP Radiation work permit SAG Severe accident guideline SAP Severe accident procedure SAR Safety analysis report NMP Unit 1 UFSAR Section I I-28 Rev. 25, October 2017 TABLE I-2 (Cont'd.) SAS Secondary alarm system SBO Station blackout SCBA Self-contained breathing apparatus SCC Stress corrosion cracking SDM Shutdown margin SDV Scram discharge volume SER Safety Evaluation Report SFC Spent fuel pool cooling and cleanup SIL Service Information Letter SJAE Steam jet air ejector SLMCPR Safety limit minimum critical power ratio SM Shift Manager SOE Sequence of events SOP Special operating procedure SORC Station Operations Review Committee SOV Solenoid-operated valve SPDS Safety parameter display system SR Surveillance requirement SRAB Safety Review and Audit Board SRLR Supplemental Reload Licensing Report SRM Source range monitor SRO Senior Reactor Operator SRP Standard Review Plan SRV Safety/relief valve SRVDL Safety/relief valve discharge line SSA Safe Shutdown Analysis SSC Structures, systems and components SWEC Stone & Webster Engineering Corporation SWP Service water system TAF Top of active fuel TBCLCW Turbine building closed loop cooling water TCV Turbine control valve TDH Total developed head TEDE Total effective dose equivalent TER Technical Evaluation Report TIP Traversing in-core probe TLAA Time-Limited Aging Analyses TLD Thermoluminescence dosimeter TMI Three Mile Island TSC Technical Support Center TSVC Turbine stop valve closure TVD Test, vent and drain NMP Unit 1 UFSAR Section I I-29 Rev. 25, October 2017 TABLE I-2 (Cont'd.) UBC Uniform Building Code UFSAR Updated Final Safety Analysis Report UHS Ultimate heat sink UL Underwriters' Laboratories Inc. Unit 1 Nine Mile Point Nuclear Station - Unit 1 Unit 2 Nine Mile Point Nuclear Station - Unit 2 UPS Uninterruptible power supply URC Ultrasonic resin cleaning U.S. United States USBM U.S. Bureau of Mines USE Upper-shelf energy USLS U.S. Land Survey UT Ultrasonic testing VWO Valve wide open WNT Waste neutralizer tank WSLR Within scope of license renewal A-13q01 PIPING, INSTRUMENT ANO EQUIPMENT SYMBOLS ON D E -RC -IC -c -cv C CRC CIC Po R RCY TEMPERATURE T TRC TIC TC TCV *s VcVcRC VclC VrC VI s SG %RC . NIC Oz Sp SpRC ISplC SpC D DC Hz HzRC 'iz INSTRUMENT SYMBOLS x J..OCAL HOl-"'TED INST. .J21NEL HOl-"'TED INST. x J..OCAL MOUNTED TRANS. .!:.,ANEL MOUNTED TRANS. 5 I ;;:: I f .... -sv Dff -R -I -G -E -x CR Cl CE ex DPR DPI FOR Kl K2 FR Fl FG FE FX HR HI HE HX pHR pHI pHE pHX LR LI LG LE PSV POR PR Pl PX Prfl. Pol RR RI RE RX TR Tl TE TX vwi. y sv Vcl sx NR NI NE NX 0;2R Ozl OzE Sf! Spl SpE DE HzE F1zR F12E CONTROL VALVE OPERATION btl° FLOAT OPERATED H.O. -MOTOR OPERATEOIAC OR DC> A.O. -AIR PISTON OPER. H.O. -HYDRAULIC PISTON OPER. P.O. -POWER DPER. DIAPHRAGM OPERATED PRESSURE CONTROLLER !SHOW FLOW DIRECTION> .£--SERYICEISEEJABLE> v ANNUNCIATOR ALARM MJOIBLEANO/ORVISJBLE ctflctl DIFF. PRESS. CONTROLLER !SHOW FLOW DIRECTION> PtEUMATIC DIAPHRAGM OPERATED WITH POSITIONER SOLENOID VAL VE REMOTE HAii.JAL CONTROL !CONTROL ROOM> EXCEPT AS NOTED OTHERWISE REMOTE HAii.JAL CONTROL !LOCAL> I I f i -T -o -AH -SH CSH CAM I CA DPT DPA I DPAH I OPAL FT FD FAwlFAi PHl>H pHSH LT L5E LAH I LAI_ PT PSE PAM PoT I I RAH RSM -TT TAM Y9T lVeAMJ Vffo VeA VcT *cs VrA IVr'.AMl*c:AL SSH NT NO NAM NAINAwlNAilNPw SUFFIX SUB 'H' DR 'L' !HIGH, LOW! SUFFIX SUB 'D' OR 'C' !OPEN, CLOSED> PIPING SYMBOLS -z ---FLOW DEVICES 17 LOW ELEMENT ORFICE -@-l*I r\ ........ ,' LOW ELEMENT NOZZLE y r-.. .eJ PILOT TUBE -D---I PLUGGED OR PUMPS ANO FANS FLANGED ENO CAP -3 PUMP --1 RECIPROCATING 1-"'IT PUMP COMPRESSOR @FAN OR BLOWER l=I FLANGED SPOOLPIECE --ffi3--ELECTRIC HEATER 0 ROTARY 1-"'IT PUMP BLOWER COMPRESSOR 0-SPRAY NOZZLE OR MIXING JET VALVE SYMBOLS --1><)----ee::r -(:::::+-CHECK IUSUALL Y SWING TYPE> -&----&--f'-.J-BUTTERFLY !CLOSE! --1-+-I-if r + -IOI-BUTTERFLY !OPEN! SAFETY OR RELIEF THREE WAY Z PORT COCK, PLUG OR BALL VAL VE -E9-Tffif:E-PORT COCK 4--cS:>--FLOW CHECK VALVE --t-1----1!1-E1 VEN Tl LA TION DAMPERS -=. --c;:::..-BACK DRAFT -i-VALVE POSITIONS Q!!RINQ 11$11Af Pl ANT QPERATl!lN -I><)---1!1---++---1+1---1l1-TO CLOSE VALVE. ALWAYS 3 WAY SOLENOID VENT SHOWN DE-ENERGIZED EXPANSION JOINTS NOT SHOWN ON P& I WHEN USED ONLY FOR NORMAL FLEXIBILITY --0--:J-BELLOWS JOINT LINE SYMBOLS INSTRUMENT PROCESS AIR INSTRUMENT CONTROL AIR INSTRUMENT ELECT. LEADS -+ INSTRUMENT CAPILLARY TUBING .1 -T--0 --<D----(D-FLOW DIRECTION SPECIAL NOTE -I><)-L.O. -LOCKED OPEN L.C. -LOCKED CLOSED R.H. -REMOTE HANUALIHANDWHEEL EXTENSION> L.R. -LANTERN RING B.S. -BELLOW SEAL K.0. *KEY OPERATED R.P.S. -REACTOR PROTECTION SYSTEM

• SUFFIX SUB 'H' OR 'L' HAY BE ADDEO !HIGH, LOW! SUFFIX SUB *o* OR *c* HAY BE ADDED !OPEN. CLOSED! CONTROL VALVE QESIGNATION !SEE TABLE OR LIST BELOW> IV -ORYWELL ISOLATION VAL VE BV -BLOCKING VAL VE -I I-CONTROL VALVE OPERATION USE APPLICATOR OPERATOR & VALVE SYHBOL OPENS ON LOSS CLOSES ON LOSS OPENS ON LOSS CLOSES ON LOSS OF MOTIVE POWER OF MOTIVE POWER OF MOTIVE POWER OF MOTIVE POWER USE APPLICABLE OPERATOR & VALVE SYMBOL OPEN OPENS ON LOSS OF HOTIVE POWER CLOSES ON LOSS OF MOTIVE POWER FAILS AS IS ON LOSS OF MOTIVE POWER FIGURE I -1 UFSAR Rev1s1on 20 October 2007

NMP Unit 1 UFSAR Section II EF II-1 Rev. 25, October 2017 LIST OF EFFECTIVE FIGURES SECTION II Figure Number Revision Number II-1 14 II-2 14 II-3 14 II-4 14 II-5 14 II-6 14 NMP Unit 1 UFSAR Section II STATION SITE AND ENVIRONMENT TABLE OF CONTENTS Section Title Section II II-i Rev. 25, October 2017 A. SITE DESCRIPTION 1.0 General 2.0 Physical Features 3.0 Property Use and Development B. DESCRIPTION OF AREA ADJACENT TO THE SITE 1.0 General 1.1 Population 2.0 Agriculture, Industrial and Recreational Use 2.1 Agricultural Use 2.2 Industrial Use 2.2.1 Toxic Chemicals 2.3 Recreational Use C. METEOROLOGY D. LIMNOLOGY E. EARTH SCIENCES F. ENVIRONMENTAL RADIOLOGY G. REFERENCES NMP Unit 1 UFSAR Section II Section II II-ii Rev. 25, October 2017 LIST OF TABLES Table Number Title II-1 1980 POPULATION AND POPULATION DENSITY FOR TOWNS AND CITIES WITHIN 12 MILES OF NINE MILE POINT - UNIT 1 II-2 CITIES WITHIN A 50-MILE RADIUS OF THE STATION WITH POPULATIONS OVER 10,000 II-3 REGIONAL AGRICULTURAL USE II-4 REGIONAL AGRICULTURAL STATISTICS - CATTLE AND MILK PRODUCTION II-5 INDUSTRIAL FIRMS WITHIN 8 KM (5 MI) OF UNIT 1 II-6 PUBLIC UTILITIES IN OSWEGO COUNTY II-7 PUBLIC WATER SUPPLY DATA FOR LOCATIONS WITHIN AN APPROXIMATE 30-MILE RADIUS II-8 RECREATIONAL AREAS IN THE REGION II-9 SOURCES OF TOXIC CHEMICALS WITHIN 8 KM (5 MI) OF UNIT 1 SITE II-10 PREDICTED VAPOR CONCENTRATION IN THE UNIT 1 CONTROL ROOM NMP Unit 1 UFSAR Section II Section II II-iii Rev. 25, October 2017 LIST OF FIGURES Figure Number Title II-1 STATION LOCATION II-2 AREA MAP II-3 SITE TOPOGRAPHY II-4 POPULATION DISTRIBUTION WITHIN A 12 MILE RADIUS OF THE STATION II-5 COUNTIES AND TOWNS WITHIN 12 MILES OF THE STATION II-6 1980 POPULATION DISTRIBUTION WITHIN A 50 MILE RADIUS OF THE STATION NMP Unit 1 UFSAR Section II II-1 Rev. 25, October 2017 SECTION II STATION SITE AND ENVIRONMENT A. SITE DESCRIPTION 1.0 General The Nine Mile Point Nuclear Station - Unit 1 (Unit 1), owned by Nine Mile Point Nuclear Station, LLC (NMPNS), is located on the western portion of the Nine Mile Point promontory. Approximately 300 ft due east is Nine Mile Point Nuclear Station - Unit 2 (Unit 2). The eastern portion of the promontory is comprised of the James A. FitzPatrick Nuclear Power Plant. The site is on Lake Ontario in Oswego County, approximately 5 mi north-northeast of the nearest boundary of the city of Oswego. Figure II-1 shows the Station location on an outline map of the state of New York. It is 230 mi northwest of New York City, 143.5 mi east-northeast of Buffalo, and 36 mi north-northwest of Syracuse. Figure II-2 is a detailed map of the area within about 50 mi of the Station. 2.0 Physical Features Figure II-3 is a detailed site map showing Station location; an associated plot plan is presented as Figure III-1 of the following section. Station buildings are situated in the western quadrant of a 200-acre cleared area centrally located along the lakeshore. Site property consists of partially-wooded land formerly used almost exclusively for residential and recreational purposes. For many miles west, east, and south of the site the country is characterized by rolling terrain rising gently up from the lake. Grade elevation at the site is 10 ft above the record high lake level, while underlying rock structure is among the most structurally stable in the United States (U.S.) from the standpoint of tilting and folding. There is no record of wave activity, such as seiche or tsunami, of such a magnitude as to make inundation of the site likely. A shore protection dike composed of rock fill from the excavation separates the buildings and the lake. NMP Unit 1 UFSAR Section II II-2 Rev. 25, October 2017 All elevations in this report refer to the United States Land Survey (USLS) 1935 data. 1. To convert elevations to 1955 International Great Lakes Data (IGLD 1955), subtract 0.375m (1.23 ft). 2. To convert elevations to 1985 International Great Lakes Data (IGLD 1985), subtract 0.217m (0.71 ft). Exclusion distances for the site are approximately 1 mi to the east, a mile to the southwest, and over a mile to the southern site boundary. 3.0 Property Use and Development There are no residences, agricultural or industrial developments (other than the James A. FitzPatrick Nuclear Power Plant) on the site; all former summer homes and farm buildings have been removed. Site boundaries and the former country road which traverses the site are posted as private property. The area immediately around the Station buildings is fenced, with building access controlled by Station security personnel. A Corporate Training Center, and the Nuclear Learning Center are located about 1,000 ft west of the Station, per Figure II-3. These installations may be reached by the public over private drives maintained by the company. B. DESCRIPTION OF AREA ADJACENT TO THE SITE 1.0 General The Station is located on the Lake Ontario coast in the town of Scriba in the north-central portion of Oswego County, approximately 5 mi north-northeast of the nearest boundary of the city of Oswego. 1.1 Population Population growth in the vicinity of the Station has been very slow, with the city of Oswego showing a decrease in population. The 1960 census enumerated 22,155 residents compared to approximately 19,793 people in 1980. However, county population increased from 86,118 in 1960 to 113,901 in 1980. The total 1980 population within 12 mi of the Station is estimated to be 46,349 (see Figure II-4). This area contains all or portions of one city and ten towns. Population and population density for the ten towns and one city within this area are shown in Table NMP Unit 1 UFSAR Section II II-3 Rev. 25, October 2017 II-1. Counties and towns within this area are shown on Figure II-5. Transient population within 12 mi of the Station is limited due to the rural, undeveloped character of the area. There are, however, a number of school, industrial, and recreational facilities in the area that create small daily and seasonal changes in area populations. The population within a 50-mi area surrounding the Station was approximately 914,193 in 1980 (see Figure II-6). The city of Syracuse is the largest population center within this area, with a population of 170,105 in 1980. Table II-2 lists cities within this 50-mi radius with populations over 10,000. The 50-mi radius contains portions of three Canadian Census Divisions located in the province of Ontario: Prince Edward, Frontenac, and Addington/Lennox. The 1976 population counts totaled 22,559, 108,052, and 32,633, respectively. 2.0 Agriculture, Industrial and Recreational Use 2.1 Agricultural Use The area within a 50-mi radius of the site encompasses all or portions of ten New York counties: Cayuga, Jefferson, Lewis, Madison, Oneida, Onondaga, Ontario, Oswego, Seneca, and Wayne. Approximately 37 percent of the land within this ten-county region is used for agricultural production. Tables II-3 and II-4 present agricultural statistics for this ten-county region. 2.2 Industrial Use Several industrial establishments are located in Oswego County, with the Novelis Corporation and the Independence Generation Plant operated by Sithe Energies USA being located nearest to the Station. The lakeshore east of Oswego is the most industrially developed area near the site. The cities of Fulton and Mexico are the only other industrial sites within 15 mi of the site. Two natural gas pipelines lie within 8 km of the plant; one pipeline supplies the Independence Plant and the other supplies Indeck Energy. Both pipelines are located on the north-south and east-west transmission line corridors. The major industrial establishments in Oswego County, their locations, and their principal products are listed in Tables II-5 and II-6. NMP Unit 1 UFSAR Section II II-4 Rev. 25, October 2017 The nearest public water supply intake in Lake Ontario is located approximately 8 mi southwest of the Station location. This intake supplies the city of Oswego and Onondaga County. Data on these and other vicinity public water supplies are listed in Table II-7. Figure II-2 shows the locations of the communities listed. 2.2.1 Toxic Chemicals Potential Sources of Toxic Chemicals According to Regulatory Guide (RG) 1.78, both onsite and offsite potential toxic gas hazards must be considered. Any toxic substance stored onsite in a quantity greater than 45 kg (100 lb) must be evaluated. Offsite sources to be evaluated include stationary facilities and frequent transportation of toxic substances (truck, rail, and barge) within 8 km (5 mi) of the site. Frequent shipments are defined as exceeding 10/yr for truck shipments, 30/yr for rail shipments, and 50/yr for barge shipments. For the NMPNS site, sources of potential toxic chemical hazards include chemicals stored onsite, as well as stationary and transportation sources within 8 km of the site. Table II-9 lists the chemicals associated with each source along with their quantities and distances from the Unit 1 control room air intake. The stationary sources include the James A. FitzPatrick plant, Novelis Corporation, Oswego Wire Inc., Sithe Independence Station, and Unit 2. One transportation source of possible hazardous materials is truck traffic along Route 104, which passes within 6.2 km (3.9 mi) of the site. Another transportation source is the railroad line between Oswego and Mexico, NY. Discussions with Conrail indicate that on average, only one hazardous chemical shipment during an 18-mo period passes through the Oswego terminal. Traffic on a spur to the site is not frequent enough (<30/yr) to warrant consideration. Only those chemicals that have the potential to form a toxic vapor cloud or plume after release to the environment need to be evaluated. This criterion is met by all chemicals listed in Table II-9. Control Room Habitability Determination The effect of an accidental release of each of the chemicals described in the previous section on control room habitability is evaluated by calculating vapor concentrations inside the NMP Unit 1 UFSAR Section II II-5 Rev. 25, October 2017 control room as a function of time following the accident. This calculation is performed using the conservative methodology outlined in NUREG-0570 and utilizing the assumptions described in RG 1.78. In a postulated accident, the entire content of the largest single storage container is released, resulting in a toxic vapor cloud and/or plume that is conservatively assumed to be transported by the wind directly toward the control room intake. The formation of the toxic cloud and/or plume is dependent on the characteristics of the chemical and the environment. The entire amount of a chemical stored as a gas is treated as a puff or cloud that has a finite volume determined from the quantity and density of the stored chemical. A substance stored as a liquid with a boiling point below the ambient temperature forms an instantaneous puff due to flashing (rapid gas formation) of some fraction of the stored quantity. The remaining liquid forms a puddle that quickly spreads into a thin layer on the ground, subsequently vaporizing and forming a ground-level vapor plume. A high boiling point liquid (above ambient temperature) forms a puddle that evaporates by forced convection with no flashing involved. The calculations are done by a computer program (VAPOR) or a spreadsheet, both based on NUREG-0570 methodology that requires the following input information: chemical physical properties, control room parameters, meteorology, distance from the spill to the control room intake, and the quantity of chemical released. The following Unit 1 control parameters are used: ventilation rate of 2530 ft3/min, and net free volume of 130,600 ft3. The most conservative meteorological conditions are assumed for the calculations, consisting of Pasquill Class A stability, 0.5 m/sec wind speed and an ambient temperature of 33°C for sodium bisulfite solution stored onsite, and Class F stability, a wind speed of 1.0 m/sec, and an ambient temperature of 90°F for all other chemical releases. The criteria for determining chemical toxicity and setting limits for habitability determinations are taken from regulatory guidance documents. According to RG 1.78, the toxicity limit of a chemical is the maximum concentration that can be tolerated by an average human for 2 min without physical incapacitation (severe coughing, eye burn, severe skin irritation). Standard Review Plan (SRP) Section 6.4 states that acute effects should be reversible within a short period of time (several minutes) without the benefit of medication other than the use of self-contained breathing apparatus (SCBA). The acute toxicity NMP Unit 1 UFSAR Section II II-6 Rev. 25, October 2017 limits listed in RG 1.78 are used in this study except that, where more appropriate, documented sources are available(2-5). Nonguideline toxicity limits are based on concentrations that produce no effects or minor irritation affecting mental alertness and physical coordination, assuming a 15-min exposure time. In cases where appropriate human data are not available, data are used by applying a conservative factor of 10 to lower the acute exposure limit. The effect of the continuous outside venting of the onsite sodium bisulfite storage tank on control room habitability is evaluated by calculating the maximum sulfur dioxide vapor concentration at the control room intake. The evaluation is performed using the guidance described in RG 1.78. The toxicity limit is set at the TLV-TWA limit established in NUREG/CR-5669(5) for sulfur dioxide. Results and Conclusions The results of the analysis are summarized in Table II-10, which indicates that none of the toxic chemicals evaluated have the potential to incapacitate the control room operators. 2.3 Recreational Use Seventeen state parks and one national wildlife refuge are located within a 50-mi radius of the Station. Table II-8 identifies the state parks and their facilities, capacities, and visitor counts. The Montezuma National Wildlife Refuge is located north of Cayuga Lake in Seneca County, approximately 44 mi southwest of the Station. C. METEOROLOGY An original 2-yr study was performed to determine the site meteorological characteristics. This study is presented in Section XVII-A. The meteorological monitoring system measures parameters to provide data that are representative of atmospheric conditions that exist at all gaseous effluent release points. Meteorological data is compiled for quarterly periods in accordance with the Offsite Dose Calculation Manual. This data is used to provide information which may be used to develop atmospheric diffusion parameters to estimate potential radiation doses to the public resulting from actual routine or accidental releases of radioactive materials to the atmosphere. NMP Unit 1 UFSAR Section II II-7 Rev. 25, October 2017 D. LIMNOLOGY A comprehensive research program, designed to monitor various parameters of the aquatic environment in the vicinity of Nine Mile Point, was begun in 1963. This detailed lake program was continued through 1978. Currently, an aquatic ecology study program (closely coordinated with James A. FitzPatrick Nuclear Power Plant) is conducted in the vicinity of Nine Mile Point on Lake Ontario to monitor the effects of plant operation with respect to selected ecological parameters, and to perform impingement studies on the traveling screens in the intake screenwell. This program is carried out and results reported in accordance with the station State Pollutant Discharge Elimination System (SPDES) Discharge Permit. E. EARTH SCIENCES A preconstruction evaluation of the geology, hydrology, and seismology of the Nine Mile Point promontory is presented in Section XVII-C. Subsequent inspection of rock exposed during excavations for the reactor and cooling water tunnels allowed for a more detailed study of subsurface conditions. No faults were encountered and no unusual conditions were observed. The structures rest on a firm, almost impervious rock foundation. Station seismic design criteria were based upon a conservative evaluation of the maximum earthquake ground motion which might conceivably occur at the site. This condition was calculated by assuming that the worst shock ever observed within an effective range of the site might be located at the closest position to the site at which an earthquake of any intensity occurred. The "maximum possible" shock assumed for Station structure acceleration calculations is of magnitude 7 at a 50-mi epicentral distance. Dames and Moore estimates that this shock will probably never occur unless unusual regional geologic changes take place. F. ENVIRONMENTAL RADIOLOGY Controlled releases of radioactive materials in liquid and gaseous effluents to the environment is part of normal Station operation. A Radiological Environmental Monitoring Program ensures that the release rates for all effluents are within the NMP Unit 1 UFSAR Section II II-8 Rev. 25, October 2017 limits specified in 10CFR20 and the release of radioactive material above background to unrestricted areas conforms with Appendix I to 10CFR50. Comprehensive studies were originally conducted to establish the effluent emission rates which would produce the above limiting conditions in the uncontrolled environment. Currently, a Radiological Environmental Monitoring Program(1), inclusive of Unit 1, is in operation. This program details the design objectives for control of liquid and gaseous wastes, including specifications for liquid and gaseous waste effluents, and specifications for liquid and gaseous waste sampling and monitoring. An annual Environmental Operating Report and Radioactive Effluent Release Report are prepared and submitted in accordance with the reporting requirements in the Technical Specifications. G. REFERENCES 1. Nine Mile Point Nuclear Station "Offsite Dose Calculation Manual." 2. Sax, N. E. Dangerous Properties of Industrial Materials, 3rd Edition, Van Nostrand Reinhold, New York, NY, 1968. 3. NUREG/CR-5669, "Evaluation of Exposure Limits to Toxic Gases for Nuclear Reactor Control Room Operators," July 1991. 4. CRC Handbook of Chemistry and Physics, 76th Edition, David R. Lide, Editor-in-Chief. 5. Air Contaminants - Permissible Exposure Limits, Title 29 Code of Federal Regulations Part 1910-1000, OSHA 3112, 1989. NMP Unit 1 UFSAR Section II II-9 Rev. 25, October 2017 TABLE II-1 1980 POPULATION AND POPULATION DENSITY FOR TOWNS AND CITIES WITHIN 12 MILES OF NINE MILE POINT - UNIT 1 Population Density 1980 Population (People Per Square Mile) City of Oswego 19,793 2665.2 Oswego (town) 7,865 302.7 Granby 6,341 142.9 Richland 5,594 105.9 Scriba 5,455 137.0 Volney 5,358 119.1 Mexico 4,790 108.3 Hannibal 4,027 99.7 Palermo 3,253 81.8 New Haven 2,421 82.1 Minetto 1,905 325.0 NMP Unit 1 UFSAR Section II II-10 Rev. 25, October 2017 TABLE II-2 CITIES WITHIN A 50-MILE RADIUS OF THE STATION WITH POPULATIONS OVER 10,000 Population City County (1980 Census) Newark Village Wayne 10,017 Clay Onondaga 52,838 Cicero Onondaga 23,689 Manlius Onondaga 28,489 Dewitt Onondaga 26,868 Syracuse Onondaga 170,105 Geddes Onondaga 18,528 Camillus Onondaga 24,333 Onondaga Onondaga 17,824 Van Buren Onondaga 12,585 Salina Onondaga 37,400 Fulton Oswego 13,312 Oswego Oswego 19,793 Oneida Madison 10,810 Rome Oneida 43,826 Watertown Jefferson 27,861 NMP Unit 1 UFSAR Section II II-11 Rev. 25, October 2017 TABLE II-3 REGIONAL AGRICULTURAL USE County Agricultural Use (square miles) Corn (All Purposes) (acres) Wheat (acres) Fruit (acres) Totals (acres) Cayuga Jefferson Lewis Madison Oneida Onondaga Ontario Oswego Seneca Wayne Totals 560 847 373 407 612 336 511 267 299 418 4,630 84,002 42,501 14,201 28,001 35,601 45,002 59,101 13,200 31,502 40,499 393,610 11,999 499 - 400 1,401 4,900 21,500 11,001 16,501 5,001 73,202 395 - - 173 222 1,097 2,330 845 954 25,125 31,141 96,396 43,000 14,201 28,574 37,224 50,999 82,931 25,046 48,957 70,625 497,953 SOURCE: NMP2 Environmental Report, Tables 2.2-9 and 2.2-10 NMP Unit 1 UFSAR Section II II-12 Rev. 25, October 2017 TABLE II-4 REGIONAL AGRICULTURAL STATISTICS - CATTLE AND MILK PRODUCTION All Cattle and Calves Beef Cows Milk Cows Average Milk Production/Cow (lb) Cayuga County Jefferson County Lewis County Madison County Oneida County Onondaga County Ontario County Oswego County Seneca County Wayne County Region State 51,000 84,000 59,000 60,000 65,000 32,500 33,000 25,500 11,500 19,000 440,500 1,780,000 2,200 2,600 600 1,600 2,500 2,500 1,600 2,300 1,000 1,800 18,700 85,000 25,000 44,000 32,500 35,500 33,500 17,000 11,500 11,500 4,300 8,500 223,300 912,000 12,200 11,100 12,300 11,800 11,300 13,200 11,900 11,400 11,200 10,400 11,680 11,488 SOURCES: 1. New York Crop Reporting Service, Cattle Inventory by County - 1980; Albany, NY, 1980 2. New York Crop Reporting Service, Milk Production - 1978, Albany, NY, 1979 3. New York Crop Reporting Service, New York Agricultural Statistics - 1978, Albany, NY, 1979 NMP Unit 1 UFSAR Section II II-13 Rev. 25, October 2017 TABLE II-5 INDUSTRIAL FIRMS WITHIN 8 KM (5 MI) OF UNIT 1 Distance/ Direction from Site Firm (km) Products Employment Novelis Corporation 4.5/SW Aluminum 1,000 sheet and plate James A. FitzPatrick <1/E Electrical 500 Nuclear Power Plant generation Nine Mile Point Adjacent Electrical 1,100 Unit 2 to Unit 1 generation Sithe Energies USA 3.5/SW Electrical 75 Independence generation Generation Plant Oswego Wire 7.0/SW Copper wire 40 Incorporated NOTE: For complete listing of major industries in Oswego County, reference Oswego County Industrial Directory. NMP Unit 1 UFSAR Section II II-14 Rev. 25, October 2017 TABLE II-6 PUBLIC UTILITIES IN OSWEGO COUNTY Location Service Niagara Mohawk Power Many sites Gas Corporation New York Telephone Many sites Communications Company Penn Central Railroad -- Shipping Oswego County Telephone Oswego Communications Company Alltel New York, Inc. Fulton Communications New York Power Authority Many sites Gas and Electric NMP Unit 1 UFSAR Section II II-15 Rev. 25, October 2017 TABLE II-7 PUBLIC WATER SUPPLY DATA FOR LOCATIONS WITHIN AN APPROXIMATE 30-MILE RADIUS Distance from Site (miles) Direction from Site Town Average Output (mgd) Source of Water 0-10 SW SW ESE Onondaga (County) Oswego Mexico 36 9 0.5 Lake Ontario (intake at Oswego) Lake Ontario Three wells; two 40-ft deep, one 38-ft deep 10-20 ENE SSE NE Pulaski Fulton Sandy Creek 0.3 2 0.2 Springs Twelve wells, 30- to 70-ft deep; two wells, 21-ft deep Two wells, 21-ft deep 20-30 SE ENE SSE S SW SSW SW NE SW Central Square Orwell Phoenix Baldwinsville Fairhaven Cato Wolcott Adams Red Creek 0.08 Not available 0.35 1 0.15 0.033 0.220 0.3 0.03 One well, 24-ft deep Spring Two wells; one 25-ft deep, one 45-ft deep Four wells; one 93-ft deep, three shallow wells Spring; one well, 46-ft deep Three wells; two 55-ft deep, one 70-ft deep Lake Ontario Springs Wells and springs SOURCE: Nine Mile Point Unit 2 PSAR NMP Unit 1 UFSAR Section II II-16 Rev. 25, October 2017 TABLE II-8 RECREATIONAL AREAS IN THE REGION Park Distance and Direction from Unit (miles) County Acreage Activities/Facilities Total Capacity (No. of People) Visitor Count (April 1979 - March 1980) Selkirk Shores 9.8 NE Oswego 980 Camping, picnicking, hiking, swimming 3,646 305,000 Battle Island 10.5 S Oswego 235 Golfing, fishing, hiking 303 40,000 Frenchman Island 26.7 SE Oswego 26 Fishing, hiking, picnicking, boating 100

• Fair Haven Beach 18.3 SW Cayuga 845 Camping, picnicking, boating, fishing 6,247 352,000 Southwick Beach 19.1 NE Jefferson 472 Camping, picnicking, boating, fishing, swimming, hiking 4,401 70,000 Westcott Beach 29.3 NE Jefferson 319 Camping, picnicking, boating, fishing, swimming, hiking 4,494 72,000 Long Point 36.0 NE Jefferson 23 Camping, picnicking, boating, fishing, swimming 754 9,000 Cedar Point 47.8 NE Jefferson 48 Camping, picnicking, boating, fishing, swimming 1,853 60,000 Burnham Point 45.4 NE Jefferson 12 Camping, picnicking, boating, fishing, swimming 553 15,000 Whetstone Gulf 48.0 ENE Lewis 2,000 Camping, picnicking, swimming, hiking 1,981 28,000 Chittenango Falls 47.2 ENE Madison 183 Camping, picnicking, hiking 699 115,000 Verona Beach 41.9 SE Madison 1,735 Picnicking, swimming 4,374 305,000 Lock 23 - Brewerton 21.6 SSE Onondaga
• Picnicking, boating 119
• Green Lakes 38.7 SSE Onondaga 1,101 Camping, picnicking, hiking, boating, fishing, swimming 3,361 1,015,000 Clark Reservation 39.1 SSE Onondaga 290 Picnicking, hiking, playground 1,255 356,000 Cayuga Lake 45.7 SSW Seneca 135 Camping, picnicking, swimming, boating, playground 3,270 129,000 NMP Unit 1 UFSAR Section II II-17 Rev. 25, October 2017 TABLE II-8 (Cont'd.) Park Distance and Direction from Unit (miles) County Acreage Activities/Facilities Total Capacity (No. of People) Visitor Count (April 1979 - March 1980) Chimney Bluffs 30.8 WSW Wayne 597 Camping, picnicking, swimming, boating, playground 1,036 30,000 NOTE: All facilities are seasonal (summer).
• Not available

, / STATION LOCATION *UTICA ALBANY* AOUM 11-1 Uf&AR Revision 14 (June 11Ht ( AREA MAP / RGURE H-2 UFIAR Reviaion 14 (June 1996) ( ., '" \,. 400' o* soo* t6oo* SITE TOPOGRAPHY FIGURE H-3 UFSAR Revision 14 (June 1996> POPULATION DISTRIBUTION WITHIN A TWELVE MILE RADIUS OF THE STATION L A K £ ONTARIO L.. -.. \(-A, Fulton LAKE(._) ' NEATAHWANTA ! ).... *-** -** -.\****** ., ' 0 : .. '**** / 12MI / .:**, ' Central Square ..... Cicero SCALE-MILES FIG.URE H-4 ' Greenboro ,, ..

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( .. -, i I I J 0 N 1980 POPULATION DISTRIBUTION WITHIN A 50 MILE RADIUS OF THE STATION <t\ ., J t I \ .,._,," '* / ,,.. ........ // ./ : .. * *., ... , *.* I I 'f66,6eJ L.--, p ( , . .r, .. .... . ,. -t.iro;i89)11 \ ... * .-.+.;--------' ------. r---\ flS0,49!1) I 0 10 20 SCALE-MILES RGURE U-6 UfSAR Revision 14 (June 1996)

NMP Unit 1 UFSAR Section III EF III-1 Rev. 25, October 2017 LIST OF EFECTIVE FIGURES SECTION III Figure Number Revision Number III-1 24 III-2 16 III-3 20 III-4 20 III-5 20 III-6 16 III-7 20 III-8 16 III-9 16 III-10 14 III-11 14 III-12 17 III-13 24 III-14 24 III-15 17 III-21 18 III-22 19 III-23 20 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section III III-i Rev. 25, October 2017 SECTION III BUILDINGS AND STRUCTURES A. TURBINE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating and Ventilation System 2.3 Smoke and Heat Removal 2.4 Shielding and Access Control 2.5 Additional Building Cooling 3.0 Safety Analysis B. CONTROL ROOM 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating, Ventilation and Air Conditioning System 2.3 Smoke and Heat Removal 2.4 Shielding and Access Control 3.0 Safety Analysis C. WASTE DISPOSAL BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating and Ventilation System 2.3 Shielding and Access Control NMP Unit 1 UFSAR Section Title Section III III-ii Rev. 25, October 2017 3.0 Safety Analysis D. OFFGAS BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating and Ventilation 1.5 Shielding and Access Control 2.0 Structure Design 2.1 General Structural Features 2.2 Heating and Ventilation System 2.3 Shielding and Access Control 3.0 Safety Analysis E. NONCONTROLLED BUILDINGS 1.0 Administration Building 1.1 Design Bases 1.1.1 Wind and Snow Loadings 1.1.2 Pressure Relief Design 1.1.3 Seismic Design and Internal Loadings 1.1.4 Heating, Cooling and Ventilation 1.1.5 Shielding and Access Control 1.2 Structure Design 1.2.1 General Structural Features 1.2.2 Heating, Ventilation and Air Conditioning 1.2.3 Access Control 1.3 Safety Analysis 2.0 Sewage Treatment Building 2.1 Design Bases 2.1.1 Wind and Snow Loadings 2.1.2 Pressure Relief Design 2.1.3 Seismic Design and Internal Loadings 2.1.4 Electrical Design 2.1.5 Fire and Explosive Gas Detection 2.1.6 Heating and Ventilation 2.1.7 Shielding and Access Control 2.2 Structure Design 2.2.1 General Structural Features 2.2.2 Ventilation System 2.2.3 Access Control 3.0 Energy Information Center 3.1 Design Bases NMP Unit 1 UFSAR Section Title Section III III-iii Rev. 25, October 2017 3.1.1 Wind and Snow Loadings 3.1.2 Pressure Relief Design 3.1.3 Seismic Design and Internal Loadings 3.1.4 Heating and Ventilation 3.1.5 Shielding and Access Control 3.2 Structure Design 3.2.1 General Structural Features 3.2.2 Heating and Ventilation System 3.2.3 Access Control F. SCREENHOUSE, INTAKE AND DISCHARGE TUNNELS 1.0 Screenhouse 1.1 Design Basis 1.1.1 Wind and Snow Loadings 1.1.2 Pressure Relief Design 1.1.3 Seismic Design and Internal Loadings 1.1.4 Heating and Ventilation 1.1.5 Shielding and Access Control 1.2 Structure Design 2.0 Intake and Discharge Tunnels 2.1 Design Bases 2.2 Structure Design 3.0 Safety Analysis G. STACK 1.0 Design Bases 1.1 General 1.2 Wind Loading 1.3 Seismic Design 1.4 Shielding and Access Control 2.0 Structure Design 3.0 Safety Analysis 3.1 Radiology 3.2 Stack Failure Analysis 3.2.1 Reactor Building 3.2.2 Diesel Generator Building 3.2.3 Screen and Pump House H. SECURITY BUILDING WEST AND SECURITY BUILDING ANNEX NMP Unit 1 UFSAR Section Title Section III III-iv Rev. 25, October 2017 I. RADWASTE SOLIDIFICATION AND STORAGE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design and Internal Loadings 1.4 Heating, Ventilation and Air Conditioning 1.5 Shielding and Access Control 2.0 Structure and Design 2.1 General Structural Features 2.2 Heating, Ventilation and Air Conditioning 2.3 Shielding and Access Control 3.0 Use J. COLD STORAGE BUILDING K. REFERENCES NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section III III-v Rev. 25, October 2017 III-1 PLOT PLAN III-2 STATION FLOOR PLAN - ELEVATION 225-6 III-3 STATION FLOOR PLAN - ELEVATIONS 237-0 AND 250-0 III-4 STATION FLOOR PLAN - ELEVATION 261-0 III-5 STATION FLOOR PLAN - ELEVATIONS 277-0 AND 281-0 III-6 STATION FLOOR PLAN - ELEVATIONS 281-0 AND 291-0 III-7 STATION FLOOR PLAN - ELEVATIONS 298-0 AND 300-0 III-8 STATION FLOOR PLAN - ELEVATIONS 317-6 AND 318-0 III-9 STATION FLOOR PLAN - ELEVATIONS 320-0, 333-8, 340-0 AND 369-0 III-10 SECTION BETWEEN COLUMN ROWS 7 AND 8 III-11 SECTION BETWEEN COLUMN ROWS 12 AND 14 III-12 TURBINE BUILDING VENTILATION SYSTEM III-13 LABORATORY AND RADIATION PROTECTION FACILITY VENTILATION SYSTEM III-14 CONTROL ROOM VENTILATION SYSTEM III-15 WASTE DISPOSAL BUILDING VENTILATION SYSTEM III-16 WASTE DISPOSAL BUILDING EXTENSION VENTILATION SYSTEM III-17 OFFGAS BUILDING VENTILATION SYSTEM III-18 TECHNICAL SUPPORT CENTER VENTILATION SYSTEM III-19 CIRCULATING WATER CHANNELS UNDER SCREEN AND PUMP HOUSE - NORMAL OPERATION NMP Unit 1 UFSAR Figure Number Title Section III III-vi Rev. 25, October 2017 III-20 CIRCULATING WATER CHANNELS UNDER SCREEN AND PUMP HOUSE - SPECIAL OPERATIONS III-21 INTAKE AND DISCHARGE TUNNELS PLAN AND PROFILE III-22 STACK - PLAN AND ELEVATION III-23 STACK FAILURE - CRITICAL DIRECTIONS NMP Unit 1 UFSAR Section III III-1 Rev. 25, October 2017 SECTION III BUILDINGS AND STRUCTURES The structural design of buildings and components is based on the maximum credible earthquake motion outlined in Volume II of the Preliminary Hazards Summary Report (PHSR). Specifically, this maximum motion consists of a magnitude 7 (Intensity IX) shock at an epicentral distance of 50 mi from the site. The maximum ground motion acceleration is 11 percent of gravity and the maximum response acceleration is 45 percent of gravity for oscillations in the period range of 0.2 to 0.3 sec. All critical structures for the Station were subjected to a dynamic response analysis for the determination of maximum stresses in the structure. Class I structures and components whose failure could cause significant release of radioactivity, or which are vital to safe shutdown and isolation of the reactor, were designed so that the probability of failure would approach zero when subjected to the maximum credible earthquake motion. (Acceleration response spectrum, Plate C-22, Section III, First Supplement to the PHSR.) Functional load stresses resulting from normal operation when combined with stresses due to earthquake accelerations are within the established working* stresses for the material involved in the structure or component. Primary load stresses, when combined with stresses due to temperature and pressure, together with stresses due to earthquake accelerations, are within applicable code or working* values. Class II structures and components were designed for stresses within the applicable codes relating to these structures and components when subjected to functional or operating loads. Stresses resulting from the combination of operating loads and earthquake loads or wind loads have been limited to stresses 33 1/3 percent above working* stresses in accordance with applicable codes. Class III structures and components are those of a service nature not essential for safe reactor shutdown and isolation, and failure of which would not result in significant release of radioactive materials. These structures were designed on the basis of applicable building codes with seismic and wind requirements. NMP Unit 1 UFSAR Section III III-2 Rev. 25, October 2017All major components in the Station were classified as above and analyzed to the appropriate degree. Vital fluid containers were analyzed and designed for hydrodynamic pressures resulting from earthquake motion. As a result of deflection determinations,

• Also see Section XVI, Subsection G. provisions were made for relative motion between adjacent components and structures where damage might result from differential movement and impact stresses. A list of the structures and components reviewed for seismic design is contained on pages III-1, III-2 and III-3 of the First Supplement to the PHSR. Stresses in the various structural members were investigated after the earthquake analysis was completed to verify that stresses are in compliance with those specified in the conventional codes such as those of the American Institute of Steel Construction, American Concrete Institute, and other applicable codes such as the New York State Building Code. All major structures are founded on very substantial Oswego sandstone which exists on the site at an average of 11 ft below grade. This eliminates the potential problems of soil consolidation and differential settlement. Figure III-1 is a plot plan showing the relationship of structures. A. TURBINE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings Exterior loadings for wind, snow and ice used in the design of the turbine building meet all applicable codes as a minimum. The roof and its supporting structure are designed to withstand a loading of 40 psf of snow or ice. The walls and building structure are designed to withstand an external loading of 40 psf of surface area, which is approximately equivalent to a wind velocity of 125 mph at the 30-ft level. 1.2 Pressure Relief Design NMP Unit 1 UFSAR Section III III-3 Rev. 25, October 2017To prevent failure of the superstructure due to a steam line break, a wall area of 1900 ft2 has been attached with bolts that will fail due to an internal pressure of approximately 62 psf, thus relieving internal pressure. Wall or building structure failure would occur at an internal pressure in excess of 80 psf. Subsequent calculations were performed in accordance with the AISC Manual of Steel Construction, Load & Resistance Factor Design (LRFD), First Edition, to compute the failure load of the building superstructure, and was determined to be at least 135 psf. 1.3 Seismic Design and Internal Loadings The turbine building is designed as a Class II structure. Components are either Class II or Class I, as outlined on pages III-1, III-2 and III-3 of the First Supplement to the PHSR. An analysis of the turbine building resulted in the use of the following earthquake design coefficients for the major components. Component Percent Gravity Comment Feedwater heaters 16.0 - 20.5 (calculation Based on and drain cooler used: 20.0 horizontal specific support structures 10.0 vertical) dynamic analysis Turbine generator 23.4 N-S horizontal Based on foundation 26.7 E-W horizontal specific dynamic analysis Condenser support 11.0 horizontal Based on structure 5.5 vertical specific dynamic analysis For the following components, percent gravity was 20.0 horizontal and 10.0 vertical, based on the Uniform Building Code (UBC). Steel structure supporting emergency Class I condenser makeup water storage tanks and demineralized water storage tank, condensate filters (CFS), backwash receiving tanks (BWRT), and condensate NMP Unit 1 UFSAR Section III III-4 Rev. 25, October 2017 demineralizer (CND) Motor generator (MG) sets for reactor Class II recirculating pump motors 150/35-ton overhead traveling crane Class II Structural anchors supporting main Class I steam, offgas, etc., piping Anchor bolts and associated bases and Classes I frame for support of all tanks, & II filters and pumps as well as electrical equipment. (Power boards, control consoles, etc.) Supports for moisture separators and Class II reheaters Stresses resulting from the functional or operating loads are within applicable codes relating to these structures and components. Stresses resulting from the combination of operating loads and earthquake or wind loads have been limited in accordance with applicable codes to a 33 1/3-percent increase in allowable stresses*. The adjoining walls of the turbine and reactor building superstructures are structurally separated to provide for dissimilar deformations due to earthquake motion. 1.4 Heating and Ventilation Heating and ventilation is provided for equipment protection, personnel comfort and for controlling possible radioactivity release to the atmosphere. 1.5 Shielding and Access Control Shielding is provided around much of the equipment to limit dose rates, as described in Section XII. Normal access to the turbine building is provided through the administration building. 2.0 Structure Design Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section III III-5 Rev. 25, October 2017The turbine building houses the power generation and allied equipment. The equipment arrangement and principal dimensions are shown on Figures III-2 through III-11. 2.1 General Structural Features The poured-in-place reinforced concrete building substructure and turbine generator foundation are founded on firm Oswego sandstone 15 ft to 25 ft below grade. The maximum bearing pressure on the rock, as recommended by consultants, is 40 tons/sq ft. This results in a safety factor of 18 based on actual unconfined compressive strength tests on selected specimens of rock core extracted from test borings. Some of the actual bearing pressures on the confine rock are as follows. Maximum Rock Structure Bearing Pressure Building column piers 27 tons/sq ft Crane column piers 20 tons/sq ft Walls below grade 13 tons/sq ft Turbine generator 24 tons/sq ft foundation The turbine generator foundation is isolated from the floors of the building to minimize transmission of vibration to the floors. This foundation is designed for stability under all conditions of loading, including vertical, horizontal and torque loads, and loads due to temperature changes, piping and seismic forces. Elastic deflection and vertical shortening of members and stresses resulting from such loading were taken into consideration. The turbine building superstructure consists of an enclosed structural steel frame. The lower 24 ft of building is covered with 8-in thick insulated precast concrete wall panels. From the 24-ft level to the roof, the building is enclosed with insulated metal wall panels made up of type FKX 16 x 16 and FKX 12 x 12 metallic-coated interior liner elements, 1 1/2-in insulation with a minimum density of 2 1/2 pcf and 16 B & S gage F-2 porcelainized aluminum exterior face sheets, all manufactured by H. H. Robertson Company. The roof is covered with metal decking, insulation, and a 4-ply tar roofing material flashed at the parapet walls. An overhead NMP Unit 1 UFSAR Section III III-6 Rev. 25, October 2017rolling door at the west end of the building provides rail car access into the building. 2.2 Heating and Ventilation System The turbine building ventilating system, shown on Figure III-12, is designed to provide filtered and heated air at an approximate rate of one change per hour, corresponding to 170,000 cfm. Two independent air supply systems are provided, each consisting of a fresh air intake, filter, electric heating unit, flow control damper, two fans, dampers and ductwork to distribute air to various areas in the turbine building. Each fan system is capable of supplying one-half of the required air, and either of the two fans in each system is considered an installed spare. The air duct electrical heating units are automatically controlled to maintain the supply air temperature at the desired level. The exhaust air system consists of two full-capacity fans, with one fan considered an installed spare, and connecting ductwork designed to induce flow of air through areas of progressively higher contamination potential prior to final discharge to the stack. An air inlet is located in each room and at each piece of equipment or other place where radioactive contamination in the form of dust, gas or vapor could be released. Ducts from these areas lead to an exhaust air manifold with each duct having a manually set control damper. The radiation protection and laboratory facilities ventilating system, shown on Figure III-13, discharges directly to the turbine building exhaust duct. In case power to the turbine building ventilation system is lost, an alternate outside source of filtered and heated air is available to the laboratory area. This area includes the technician's office, auxiliary calibration laboratory, high level lab, low level lab, counting room, auxiliary counting room and instrument calibration room. A shunt circuit draws air from the exhaust manifold and monitors its airborne radioactivity. The circuit is located so that it monitors building air conditions and not the exhaust from equipment vents. High activity causes alarm in the Station control room. NMP Unit 1 UFSAR Section III III-7 Rev. 25, October 2017The exhaust system discharges into the plenum which also receives air from the containment and other buildings, as shown on Figure VI-24. Backflow from other systems to the turbine building is prevented by interlocks which require valves to be closed if the exhaust fans are not in operation. The turbine building ventilating system is designed to maintain the building at a slightly negative pressure. Differential pressure control will automatically regulate air supply to maintain a negative pressure within the building with respect to the outside. This is to control the release of contaminated air and prevent out-leakage. Electrical heaters are provided in various areas of the building for auxiliary heat should the ventilation system not be in operation for any reason. Water-cooled heat exchanger cooling units are provided in areas surrounding the extraction heaters, moisture separators, condensate circulating pumps and reheaters to dissipate the radiant heat loss from this equipment and to maintain desired temperatures for personnel comfort and equipment protection. The cooling water is supplied from the service water system. 2.3 Smoke and Heat Removal Smoke and heat removal capability is provided for the three smoke zones on el 250' of the turbine building and the upper elevation of the turbine building. Twelve motor-operated vents are installed in the roof over the turbine generator, and five sidewall vents are installed in the wall at el 351'. A fire which produces low heat but a large concentration of smoke will be vented through the roof and sidewall vents. This capability is provided by manual actuation of the motor-operated vents. High heat and high smoke fires will automatically open the roof vents when the fusible link trips. In addition, the railroad access door on el 261' will be remotely opened to assist in smoke purging. 2.4 Shielding and Access Control Personnel access into the turbine building is controlled from the administration building at el 248'-0". An elevator for operating personnel serves the entire seven floor levels in the turbine building and is located at H row between column lines 11 and 12 (Figures III-4 through III-9). NMP Unit 1 UFSAR Section III III-8 Rev. 25, October 2017Stairs are also provided alongside the personnel elevator to serve the seven floor levels. In addition to the main or full-height stairs, stairs are provided at four locations at grade for accessibility to floors above grade, and at seven locations to serve floors below at el 250' and el 237'. Walls, floors and roofs around equipment containing radioactivity are designed to have concrete thicknesses which significantly reduce radiation levels, as discussed in Section XII. 2.5 Additional Building Cooling During warm weather conditions, the turbine building roof vents and/or exterior doors may be opened to provide additional building cooling. When the roof vents or doors are open, the turbine building differential pressure may approach zero in localized areas. In such cases, procedural controls for air sampling are used to prevent an unmonitored release of radioactivity to the environment. 3.0 Safety Analysis The turbine building walls are of noncombustible material consisting of poured-in-place concrete, precast concrete, or insulated metal panels. The turbine room internal roof also consists of noncombustible material. Metal decking spans the steel purlins and is covered with rigid insulation and 4-ply built-up roofing material. All floors are of noncombustible material: either poured concrete or steel grating. Pressure relief to prevent failure of the superstructure due to a steam line break has been provided in the metal wall siding on the north wall of the crane bay (column Row C). A peripheral drain at the exterior of the building provides for the removal of groundwater seepage and discharges into a sump pit with pump at the low point of all the buildings (southwest exterior corner of the reactor building). A rock dike 1000-ft long at the shoreline protects the Station from lake wave action or possible ice accumulation. The dike is 2 ft higher than yard grade and is constructed of rock from the Station excavation. Large rocks face the lake side of the dike and have proven very effective in wave damping and as a barrier to floating ice. NMP Unit 1 UFSAR Section III III-9 Rev. 25, October 2017The turbine building grade floor at el 261 is 12 ft above maximum lake level (el 249). Poured-in-place concrete foundations enclose the turbine building below grade floor level, and preformed rubber water stops are incorporated in the concrete construction joints for water-tightness. B. CONTROL ROOM The control room is located in the southeast corner of the turbine building at el 277. It is bounded by the administration building offices on the south and east, the turbine room on the west, and the control room break area, instrumentation and control (I&C) office area, and diesel building on the north. 1.0 Design Bases 1.1 Wind and Snow Loadings The wind and snow loadings for the control room are the same as for the turbine building. 1.2 Pressure Relief Design There are no special pressure relief requirements for the control room. 1.3 Seismic Design and Internal Loadings The structural design for the control room, as well as the auxiliary control room below at el 261, is Class I seismic based on the maximum credible earthquake motion outlined in the introduction to Section III. Components whose functional failure could cause significant release of radioactivity, or which are vital to safe shutdown and isolation of the reactor, are also designed as Class I. The seismic analysis resulted in the application of acceleration factors of 20.0 percent gravity horizontal and 10.0 percent gravity vertical. These acceleration factors were calculated from the dynamic analysis of the turbine building. Although the control room is structurally a part of the turbine building, functional load stresses when combined with stresses due to earthquake loading are maintained within the established working stresses* for the structural material involved. Also see Section XVI, Subsection G. NMP Unit 1 UFSAR Section III III-10 Rev. 25, October 20171.4 Heating and Ventilation Heating and air conditioning are provided for personnel comfort and instrument protection. The ventilating system also provides clean air to the control room following an accident. 1.5 Shielding and Access Control Normal access to the control room is provided from the administration building through security-controlled doors. Shielding is supplied to allow continuous occupancy during any reactor accident. The most limiting accidents are the control rod drop accident (CRDA) and the loss-of-coolant accident (LOCA) without core spray, which are described in Section XV. The shielding also meets the design TEDE dose rate for personnel in the control room such that the exposure limits of 10CFR50.67 will not be exceeded in the course of the LOCA. In addition, the cumulative dose from any design basis accident (DBA) would also meet 10CFR50.67 limits. Credit is taken for automatic initiation of the control room air treatment system for the LOCA. If air outside the building is contaminated, the ventilating system will be controlled to assure that contamination within the control room is minimized and kept within the above limits, as shown in Section 3.0, following. 2.0 Structure Design Plans showing location and principal dimensions are shown on Figures III-4, III-5, and III-6. 2.1 General Structural Features The structural steel enclosing the control room and the auxiliary control room below is supported on concrete walls and concrete foundations bearing on and keyed into sound rock. Actual rock bearing pressures are less than one-third of the allowable working bearing pressure. Lateral earthquake forces or wind loads are transmitted to the concrete foundations by the combination of structural steel bracing and concrete walls. The control room walls, roof and floors are framed with structural steel. The west and north interior walls are 12-in solid reinforced concrete. The east wall is enclosed with insulated metal wall panels made up of FK-16 x 16 metallic-coated interior liner elements, 1 1/2-in insulation and NMP Unit 1 UFSAR Section III III-11 Rev. 25, October 201716 B & S gage F-2 porcelainized aluminum exterior face sheets, as manufactured by H. H. Robertson Company. The wall panel joints are sealed with a synthetic elastomer caulking material. This wall is separated from the administration building extension by a 3-in rattle space. The south interior wall consists of 8-in concrete blocks laid with steel-reinforced mortar joints. An interior metal partition wall parallel to the south wall forms a 6'-6" corridor and is provided with windows for observing the control room operations from the corridor. The slab immediately above the control room at el 300 is pinned to the walls and provides radiation shielding, and consists of 8 1/2-in thick poured-in-place reinforced concrete supported on structural steel beam framing. Two-thirds of this slab area has a roof above at el 333 which is made up of 3-in deep metal decking, 2 in of insulation and a 5-ply roof with slag surface. The remaining third of the slab area provides a shielding roof over the control room and consists of the 8 1/2-in thick poured-in-place reinforced concrete slab to which is applied 1 1/2 in of rigid insulation and a 5-ply roof with slag surface. The control room floor is poured-in-place reinforced concrete on 14-gauge metal decking. The gross depth of the floor slab is 8 in and the average depth of concrete is 5 3/4 in. 2.2 Heating, Ventilation and Air Conditioning System The ventilation system shown on Figure III-14 is designed to provide outside and recirculated air to the control room and auxiliary control room areas during normal and emergency conditions. In the normal ventilation mode, outside air enters the system through a louvered intake after which it passes through a 15-kW duct heater and normal supply isolation dampers, which are interlocked with the emergency ventilation inlet dampers. Outside air is needed to recoup air from leakage and losses and to maintain a habitable environment for personnel. The outside air then flows through an outside air mix damper and is then mixed with recirculated control room return air from the recirculation damper, which is set to maintain a positive pressure in the control room. The total amount of air (14,500 cfm minimum) then passes through a two-element dust filter and redundant cooling coils where it will be cooled, if necessary, to ensure the control room temperature does not exceed the maximum calculated temperature of 80.5°F. The cooled air enters the control room circulation fan for distribution to various NMP Unit 1 UFSAR Section III III-12 Rev. 25, October 2017areas through ducts. Air will circulate through the control room to the return ductwork for recirculation and mixing with additional outside air. In order to prevent infiltration of potentially contaminated air, doors are weather-stripped and penetrations are sealed to maintain a positive pressure to the turbine building of 1/16 in of water. The emergency ventilation system is automatically initiated on high radiation signal from the intake radiation monitors, LOCA and/or MSLB signal from the reactor protection system (RPS), or manually initiated when required by procedures. The normal supply isolation dampers will be automatically closed, and the emergency ventilation inlet dampers will be opened. The outside air will then flow through a 15-kW duct heater and one of the full capacity control room emergency fans. The design flow rate for the control room emergency ventilation system is 2250 cfm +/-10%. Air then passes through a manual throttling damper, a high efficiency particulate filter, and an activated charcoal filter unit. This filtered air will then join the normal supply ductwork and mix with control room return air to be circulated by the normal control room circulation fan. The design flow rate for the emergency ventilation system outside air is determined as that necessary to maintain a positive pressure of 1/16 in of water to the turbine building, administration building, and outside atmosphere, and is a function of control room boundary leakage. The design flow rate of 2250 cfm +/-10% is within the required range of 1000 to 3750 cfm which is based on minimum required fresh air for personnel and maximum filter capability. The emergency ventilation fans may be manually started for periodic testing. Heating is provided by thermostatically-controlled ventilation duct heaters. Cooling is provided by two chiller units. Both the temperature control valve and/or the bypass valve for the chilled water system may be open without overcooling the control room. Tests and inspections on the control room emergency ventilation filters are done in accordance with Technical Specifications. 2.3 Smoke and Heat Removal To assist in maintaining a habitable atmosphere in the control room and auxiliary control room, a smoke purge capability is NMP Unit 1 UFSAR Section III III-13 Rev. 25, October 2017provided from two independent fans, one 6000-cfm makeup fan and one 8000-cfm exhaust fan (Figure III-14). Qualitative smoke evaluations have been performed for NMP1. The evaluations assessed the effects of both external and internal fire/smoke events on the capability to maintain reactor control from either the control room or remote shutdown panels. The evaluations considered various plant design and procedural criteria in accordance with RG 1.196, "Control Room Habitability at Light Water Nuclear Power Plants," and NEI 99-03, "Control Room Habitability Guidance," Revision 1. The evaluations confirmed that egress pathways to and including the remote shutdown panels are served by ventilation systems independent of the control room and that no single smoke/fire event could preclude use of both the control room and remote shutdown panels. 2.4 Shielding and Access Control Normal personnel access to the control room is provided by three controlled access doors all located on el 277. The north door opens into the control room break area, the south door opens into the administration building, and the west door opens into a corridor, giving access to the administration building at el 277 and also making available the stairway to el 261 of the administration building. In addition to the above, a stair is provided within the control room (northwest corner) down to the auxiliary control room on the ground floor, shown on Figure III-4. In case of a reactor accident, personnel access to or from the control room would be from the southerly extreme of all buildings and approximately 400 ft from the center of the reactor. The walls, roof and floors are designed to have concrete thicknesses which provide shielding during the design basis accident (DBA). 3.0 Safety Analysis The control room is designed for continuous occupancy by operating personnel during normal operating or accident conditions. Concrete shielding provided in the roof and floors above and in the walls facing the reactor building is more than sufficient to ensure the exposure limits of 10CFR50.67 will not be exceeded in the course of a LOCA. Maintaining positive pressure inside the control room and regulating the filtered NMP Unit 1 UFSAR Section III III-14 Rev. 25, October 2017outside air supply prevents the concentration of radioactive materials and ensures that the cumulative dose from the LOCA accident will be within the exposure limits of 10CFR50.67. In addition, supplied air respirators are available in the control room for use if necessary. Tracer gas testing is performed periodically using the constant injection method of ASTM E741-00, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution." For the constant injection method, a constant flow of tracer gas is injected into the control room envelope (CRE) until the resulting concentration reaches a steady state value. This occurs when the amount of tracer gas entering the CRE is the same as the amount leaving the CRE. By injecting the tracer gas in the outside airflow used for pressurization of the envelope, an estimate of the filtered and unfiltered airflow that provides this pressurization can be made by measuring the concentration of tracer gas in the outside airflow while at the same time measuring the steady state concentration in the CRE. A CRE habitability program has been established to ensure that CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. Both normal and emergency lighting are provided in the control room together with communications, air conditioning, ventilation, heating and sanitary plumbing facilities. If normal electric power service is not available, provision has been made to power the cooling, ventilating and heating units from the emergency diesel generators. Building components and finish materials are noncombustible and combustible materials are not stored in the control room. The minimum distance of the control room to the centerline of the reactor is 330 ft and there are no direct connections from passageways, ventilating ducts or tube connections between the reactor building and the control room. The floor of the control room is 16 ft above yard grade and 28 ft above maximum lake level (el 249). Therefore, the possibility of flooding or inundation is incredible. C. WASTE DISPOSAL BUILDING NMP Unit 1 UFSAR Section III III-15 Rev. 25, October 2017 1.0 Design Bases 1.1 Wind and Snow Loadings Wind and snow loadings for the waste disposal building are the same as for the turbine building. 1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 1.3 Seismic Design and Internal Loadings The waste disposal building and major components whose functional failure could cause significant release of radioactivity, or which are vital to safe shutdown and isolation of the reactor within are designed as Class I structures. The analysis of stress levels used the following earthquake design coefficients. Percent Gravity Horizontal Vertical Elevations 225 and 229 11.0 5.5 Elevation 236-6 11.5 5.5 Elevations 246-6, 247 12.2 5.5 and 248 Elevation 261 17.0 7.33 Elevation 277 (276-6) 30.7 7.33 Roof Elevation 289 30.7 7.73 Exterior walls of the substructure are designed for an earth pressure at any depth equal to the depth in feet times 90 psf. The exterior walls of the substructure and the base slab are designed to resist hydrostatic pressure and uplift due to exterior flooding to el 249. Except where concentrated loading due to the handling and placement of equipment requires construction of greater NMP Unit 1 UFSAR Section III III-16 Rev. 25, October 2017strength, the substructure floors are designed for dead loads plus the following: Live Loads Elevations Pounds Per Sq Ft 225 and 229 Unlimited 236-6, 237 and 248 350 241 and 247 250 The grade floor at el 261, including the concrete shielding plugs which close hatchways over equipment in the substructure, is designed for a uniform live load of 450 psf; or in the loading area a concentrated loading pattern produced by an AASHO* H20 loading, or 1000 psf, whichever requires the stronger construction. 1.4 Heating and Ventilation Heating and ventilation is provided for personnel comfort, equipment protection and for controlling possible radioactivity release to the atmosphere. 1.5 Shielding and Access Control Shielding is provided around tanks and equipment to maintain dose rates as described in Section XII. Normal access to the waste disposal building is from the turbine building. 2.0 Structure Design Floor and roof plans, exterior elevations, sections showing interior walls, and architectural details of the building are shown on Figures III-2 through III-6 and Figure III-11. 2.1 General Structural Features The poured-in-place reinforced concrete building substructure is founded on firm Oswego sandstone. American Association of State Highway Officials. NMP Unit 1 UFSAR Section III III-17 Rev. 25, October 2017The maximum bearing pressure on the rock as recommended by consultants is 40 tons/sq ft. This results in a safety factor of 18 based on actual unconfined compressive strength tests on selected specimens of rock core extracted from test borings. The building has a flat roof consisting of a cellular metal deck covered with insulation and a bitumen and felt roofing membrane. The exterior facing of the superstructure walls is of sheet metal, attached either to an exterior shielding wall or to insulated cellular sheet metal wall. The interior walls of the substructure are of cast-in-place concrete and those for the superstructure are either cast-in-place or made of concrete masonry units. With minor exceptions, all structural floors are poured-in-place concrete slabs. The superstructure frame is of fabricated steel. The north section of the basement is divided into three levels. These floors are for the storing of solid radioactive waste in metal drums until it is suitable for offsite shipment to a permanent disposal area. The intermediate level floor elevation is for the storage of evaporator bottoms and filter sludge prior to solidification. The south section of the basement provides space for the temporary storage, pumping and processing of radioactive liquid waste as described in Section XII. The loading area for receiving empty waste drums and equipment as described in Section XII is located on el 261 (Figure III-4). The designed control for spilled liquid is to allow the fluid to seek a lower level and, thus, be accommodated by the sumps which contain the fluid, and pump it directly to storage tanks. All drainage sumps have smooth linings of steel plate with all joints welded. The waste drum filling area also has a drainage gutter lined with half of a steel pipe. These designs are to facilitate cleanup by preventing contaminated liquids from permeating the concrete shell of the sump pit or gutter. 2.2 Heating and Ventilation System The heating and ventilating system, shown on Figure III-15, is designed to supply filtered and heated air at approximately 9,000 cfm and exhaust it after filtration. This corresponds to NMP Unit 1 UFSAR Section III III-18 Rev. 25, October 2017about one change of air per hour. No air is discharged from the building except through the stack. The supply fans, exhaust fans and exhaust filters are provided with full-capacity backups. Either supply fan and either exhaust fan can then be used to operate the system while the other members of the pairs are on standby. Outside air is drawn into the system through a fixed louver housed above the roof of the building and protected by bird and insect screening. The air is drawn through a filter designed to remove dust, and an electric heater of 200-kW capacity. The heater is thermostatically controlled to warm the air to maintain at least 70°F in accessible areas. Beyond the heater section the supply duct is split with each half routed through a supply fan of 9,000 cfm capacity. Each fan is isolated in its section of duct by a butterfly valve damper on both inlet and discharge sides. Beyond the fan discharge control dampers, the ducts rejoin into a common manifold from which supply ducts convey fresh air to various areas of the building. At or near the discharge point of each of these ducts, a manually set damper determines the fraction of air delivered at that particular point. The fresh air supply points are located where the rate of air contamination is lowest while the inlets to the exhaust ducts are located where the rate of contamination is likely to be the highest. An air outlet is located in each room and at each piece of equipment or other place where radioactive contamination in the form of dust, gas or vapor could be released. Ducts from these areas lead to an exhaust air manifold with each duct having a manually set control damper. A shunt circuit draws air from the exhaust manifold and monitors its airborne radioactivity. The circuit is located so that it monitors building air conditions and not the exhaust from equipment vents. High activity is alarmed in the Station main control room. Beyond this point, the exhaust duct divides into two full-sized parts, each of which contains a roughing filter followed by a high-efficiency filter and an exhaust fan as shown on Figure III-15. Butterfly valves in the ducts, before the filters, between filters and fans, and following the fans determine which of the alternate routes the exhaust will take and regulate the NMP Unit 1 UFSAR Section III III-19 Rev. 25, October 2017amount of air exhausted. From here on, the ducts are reunited and discharge to the plenum leading to the stack. Backflow from other systems is prevented by interlocks which require valves to be closed if the exhaust fans are not in operation. Each high-efficiency particulate filter in the exhaust system has a minimum removal efficiency of 99.97 percent based on the 0.3 micron "DOP" (dioctylphthalate smoke) test. Supplementing this exhauster system is a 300-cfm capacity auxiliary system, which exhausts air directly from the hydraulic baler through a roughing filter and a high-efficiency filter by means of a small exhauster fan, and discharges directly into the ventilation breaching. Also, a 500-cfm capacity auxiliary system exhausts directly from the drum filling area through a roughing filter by means of a small exhauster fan, and discharges to the exhaust duct of the building ventilating system. Equipment vents and the sample Station hood discharge directly to the exhaust duct. Supplementing the heat supplied by the main intake air heater, small heating units are provided locally to maintain desired temperatures for comfort of personnel and protection of equipment. The ventilation system for the waste building extension is shown on Figure III-16. One of two full-capacity exhaust fans draws air at a rate of 5300 cfm from the waste building through ductwork from the various equipment rooms within the waste building extension. The air that passes through the system is discharged to the stack. 2.3 Shielding and Access Control Normal personnel access to the waste disposal building is from the turbine building through the waste disposal control room. Access doors from the turbine building are also located near the baler room. Access is also available through the truck loading bay located at the northeast corner of the building. All access to the building is at grade level as shown on Figure III-4. All levels are accessible by steel stairways from the grade floor, and an emergency ladderway exit is provided for those parts of the drum storage area which are remote from the stairs. Hatches are provided for access to equipment. NMP Unit 1 UFSAR Section III III-20 Rev. 25, October 2017Concrete thicknesses for both walls and floors are established to provide the degree of radiation shielding of radioactive waste adjacent to the shielded area. The reinforced concrete substructure completely isolates the basement and serves as shielding for adjoining basement areas. Each item or group of closely associated items of equipment is housed in a separate room, surrounded by concrete shielding walls of appropriate thickness to provide adequate protection to operating personnel, as determined by the anticipated intensity of radiation and time duration of exposure. The waste disposal building control room is completely surrounded by shielding walls, and with access so arranged that the room will be accessible at all times. 3.0 Safety Analysis The design and construction of the waste building has provided for all foreseeable conditions and loads. All structural material used is noncombustible and accumulation of combustible material is carefully avoided. As outlined in the detailed description of the structure, provision has been made that, should some unforeseen condition or accident release contaminated waste, the hazard would be localized and the size of the cleanup and decontamination job restricted. All tanks are made of ductile metal and all sump pits are lined so that these containers can be subjected to substantial distortion without rupture. The two rooms for the centrifuges on the grade floor are surrounded by heavy walls which serve a dual purpose by providingboth radiation and mechanical shielding. The centrifuge is obsolete and is de-energized. The substructure is massive reinforced concrete, not subject to fracturing. D. OFFGAS BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings NMP Unit 1 UFSAR Section III III-21 Rev. 25, October 2017Exterior loadings for wind, snow and ice used in the design of the offgas building are the same as the turbine building. 1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 1.3 Seismic Design and Internal Loadings The offgas building is designed as a Class I structure. The analysis of stress levels used the following earthquake design coefficients. North-South East-West Elevation  % G  % G 289 37.2 32.0 276 19.3 24.0 261 15.2 19.0 247 13.6 16.0 236 12.0 13.0 The live load design on the ground floor and intermediate subfloors is 300 psf. 1.4 Heating and Ventilation Heating and ventilation is provided for personnel comfort. 1.5 Shielding and Access Control Shielding is provided around tanks and equipment to maintain dose rates as described in Section XII. Normal access to the offgas building is from the turbine building. 2.0 Structure Design Floor and roof plans, exterior elevations, sections showing interior walls, and architectural details of the building are shown on Figures III-2 through III-9. 2.1 General Structural Features NMP Unit 1 UFSAR Section III III-22 Rev. 25, October 2017The substructure is constructed of cast-in-place reinforced concrete and is founded on firm Oswego sandstone. The maximum bearing pressure on the rock is 20 tons/sq ft. This results in a safety factor of 18 based on actual unconfined compressive strength tests on selected specimens of rock core extracted from test borings. The building has a built-up roof consisting of a cellular metal deck covered with insulation and asbestos felt and a gravel surface. The superstructure is structural steel frame with insulated exterior metal walls. The interior walls of the substructure are of cast-in-place concrete and those for the superstructure are concrete block with a 144-pcf density for shielding. With minor exceptions, all structural floors are poured-in-place concrete slabs. The basement is divided into two levels. El 229' houses the charcoal column tank room. Located on el 232' is the chiller system compressors and deicing water buffer tank rooms. The next floor is divided into three levels. The main level el 247' houses the three chiller rooms and equipment hatch. El 244'-9" houses the two preadsorber rooms, and at el 250' is grating surrounding the charcoal tanks. Normal personnel and equipment access from the turbine building is located on el 261'. Also located on this level are equipment plugs, equipment hatch and stair openings to the levels below. 2.2 Heating and Ventilation System The heating and ventilation system is shown on Figure III-17. One of two exhaust fans, with a full capacity of 6,000 cfm, draws air at a rate of 5400 cfm from the turbine building through ductwork from the various equipment rooms within the offgas building. The air that passes through the system is discharged to the stack. 2.3 Shielding and Access Control Normal personnel access to the offgas building is from the turbine building. An access door from the waste disposal building is also provided. All access is located on grade level 261. All levels of the offgas building are accessible by steel stairways from the grade floor. Equipment plugs and hatch are provided for access to equipment. NMP Unit 1 UFSAR Section III III-23 Rev. 25, October 2017 Concrete thicknesses for both walls and floors were established to provide adequate radiation shielding consistent with as low as reasonably achievable (ALARA) criteria. 3.0 Safety Analysis The design and construction of the offgas building has provided for all foreseeable conditions and loads. All walls, floors and roof are of noncombustible materials. Equipment is housed in rooms with walls, floors and shield walls appropriately designed to provide adequate shielding to meet ALARA criteria. E. NONCONTROLLED BUILDINGS 1.0 Administration Building The administration building is a one and two-story structure adjoining the turbine building on the south and east. 1.1 Design Bases 1.1.1 Wind and Snow Loadings The wind and snow loadings for the administration building are the same as for the turbine building. 1.1.2 Pressure Relief Design There are no special pressure relief requirements for the administration building. 1.1.3 Seismic Design and Internal Loadings The administration building is designed as a Class II and III structure. The original administration building was designed as a Class III structure with no special seismic criteria. The following design live loads were used in addition to the dead loads for the original administration building. Elevation 261 Store room and shop room - 1000 psf Other Areas 150 psf NMP Unit 1 UFSAR Section III III-24 Rev. 25, October 2017 Elevation 277 Office areas, including areas for office equipment and personnel, corridors, stairways and other related areas - 125 psf The administration building extension is designed as a seismic Class II structure. A portion of the extension is located over the diesel generator rooms requiring an upgraded seismic classification. The extension is designed to accommodate the same seismic loads as the control room and diesel generator rooms. The criteria used for the administration building extension are: 1. Normal allowable stress* levels were used. (However, up to 1/3 overstress was permitted.) 2. Horizontal north-south and east-west earthquakes were not combined but were considered separately. 3. Vertical accelerations were assumed to be 1/2 of the horizontal. 4. Accelerations and deflections caused by the earthquake are: North-South East-West Elevation  % G  % G 300 34.0 30.0 277 19.0 18.0 261 13.0 13.0 250 12.0 12.0 1.1.4 Heating, Cooling and Ventilation Heating, cooling and ventilation are provided for personnel comfort. 1.1.5 Shielding and Access Control No shielding is required. Also see Section XVI, Subsection G. NMP Unit 1 UFSAR Section III III-25 Rev. 25, October 20171.2 Structure Design The administration building, shown on Figures III-3 through III-5, contains all the facilities required for administrative and technical servicing functions required of a nuclear generating station. 1.2.1 General Structural Features The administration building is a steel-framed structure with cellular metal and concrete floors and exterior walls of insulated sandwich precast concrete slabs. The exterior walls of the administration building extension are metal siding. The exterior south and west walls of the women's locker room and the foam room are masonry walls. The building has three levels. The basement (el 248) houses the onsite Technical Support Center (TSC). The TSC meets the requirements of NUREG-0578. The location of the TSC and its proximity to the control room is shown on Figures III-3 and III-5. This level is also used for storage, additional office space, radiologically-controlled access point to the turbine building, and personnel locker room. The ground floor (el 261') is divided into three parts. One of these is assigned to Station stores. The remaining two are assigned to shops. The balance of the ground floor contains an anteroom and a foyer for the stairway and elevator to the general offices on the second floor. On the upper level (el 277') are the stair, elevator lobby, restrooms, offices, conference rooms, and a satellite document control station. Document control, microfilming facilities, and the record storage facility, in accordance with ANSI N45.2.9-5(6), are located at Nine Mile Point Nuclear Station - Unit 2 (Unit 2). 1.2.2 Heating, Ventilation and Air Conditioning Ventilation for the administration building and the administration building extension is provided as follows. One self-contained rooftop air conditioning unit, one supply fan, three exhaust fans, and associated ductwork and equipment provide ventilation to the original administration building. NMP Unit 1 UFSAR Section III III-26 Rev. 25, October 2017 Five supply fans, associated ductwork and equipment supply air to the administration building extension. Individual heating and air conditioning units are provided throughout the original administration building and the administration building extension for personnel comfort. The onsite TSC, located on el 248', is provided with an air filtering system which is housed in the charcoal filter building at el 261' (see Figure III-18). 1.2.3 Access Control Normal access to the administration building is provided by two doors located on the west side of the building. Three overhead doors are located on the south side of the building to provide access to the shops and stores at the 261 ft level. 1.3 Safety Analysis No radioactivity complications exist at any of the noncontrolled buildings because radiation, contamination, and airborne radioactivity levels are administratively controlled. Appendix 10A provides details regarding the fire hazards analysis of the noncontrolled administration building. 2.0 Sewage Treatment Building The new sewage treatment facility (STF), which utilizes part of the existing STF, is located in the vicinity of railroad track spur no. 3 that was removed for construction, approximately 300 ft northwest of the turbine building and due west of the north end of the reactor building as shown on Figure III-1. The site was selected based on review of available areas outside the flood plain for a Unit 2 10,000-yr flood year flood (rain). The existing STF was modified to function as a raw sewage pump station and an equalization tank for the new STF. The control building for the new STF is located between and to the south of the circular extended aeration units. The control building houses a new laboratory, a motor control center (MCC), blower room, storage room, maintenance room and hypochlorite room, as well as an influent/effluent room. Normal access to the treatment units is from inside the control building's NMP Unit 1 UFSAR Section III III-27 Rev. 25, October 2017influent/effluent room. Maintenance and emergency access to the treatment unit may be from outside access doors on each tank. 2.1 Design Bases 2.1.1 Wind and Snow Loadings The wind and snow loading for the sewage treatment building design were determined by the requirements of ANSI A58.1 and the New York State Building Code. 2.1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 2.1.3 Seismic Design and Internal Loadings The sewage treatment building is designed as a Class III structure with no special seismic criteria. The system conforms to state regulations for sewage systems. 2.1.4 Electrical Design In certain areas of the building, electrical components are protected by explosion-proof enclosures. 2.1.5 Fire and Explosive Gas Detection Automatic fire detection equipment is provided in the STF. The fire detection equipment actuates alarms on local fire panels in the STF which informs personnel of fire location. Automatic gas detection equipment is provided for chlorine, and for methan and other explosive gases. The detection equipment actuates an alarm bell and warning lights inside and outside the STF. Both systems are provided for personnel safety and equipment protection. 2.1.6 Heating and Ventilation Heating and ventilation is provided for equipment protection and personnel comfort in accordance with the required codes. 2.1.7 Shielding and Access Control NMP Unit 1 UFSAR Section III III-28 Rev. 25, October 2017 Shielding is not required. 2.2 Structure Design 2.2.1 General Structural Features The sewage treatment plant will provide secondary treatment and disinfection for a minimum flow of 10,000 gal/day and a peak flow of 240,000 gal/day. Wastewater flows by gravity from Nine Mile Point Nuclear Station - Unit 1 (Unit 1) facilities, the Energy Information Center (EIC), the Nuclear Learning Center (NLC), and Unit 2 to the existing Unit 1 sewage treatment plant building and associated preliminary treatment facilities. After preliminary treatment, the flow is pumped to the extended aeration units. Flow through the remainder of the plant is by gravity. Discharge from the plant is through a 12-in outfall sewer to a drainage ditch leading to Lake Ontario. Flow measurement is available and is recorded on stripcharts. Raw sewage will pass through a comminutor to shred large solids. Two comminutors are provided, each capable of treating flows up to 300,000 gal/day. In the event of failure of both comminutors, a bypass hand-cleaned bar screen is provided to protect the raw sewage pumps from large solids. Raw sewage is then pumped to the new treatment facilities. Pumping after preliminary treatment minimizes the need for rock excavation for downstream treatment units. A 4-in and 6-in dual-force main is used to meet the anticipated flow range of 35,000 gal/day to 240,000 gal/day. A three-pump raw sewage station is utilized with two pumps operating and the third pump acting as an installed standby. Wastewater pumped to the new treatment facilities will enter a flow distribution structure and will be split equally by weirs to two extended aeration units. Each unit contains two equally-sized basins of 2800 cu ft, while affording maximum control and operational flexibility. At double outage design conditions, two units each with two basins of this size would provide an average hydraulic detention time of approximately 17 hr with an average organic loading of about 18 lb biological oxygen demand (BOD) per day per 1000 cu ft of tank volume. NMP Unit 1 UFSAR Section III III-29 Rev. 25, October 2017The aeration system for the activated sludge process is a coarse-bubble diffused air system. A total of three air blowers (including standby) are provided, having a total capacity of 700 scfm. These blowers will provide approximately 3200 cu ft of aeration air per pound. The mix liquor is then sent to the activated sludge settling tank where the sludge solids are separated. This produces a well-clarified effluent low in BOD and suspended solids. Each treatment unit contains an 18-ft diameter clarifier with 12-ft side water depth. These tanks are center feed clarifiers with radial outward flow. At double outage design conditions, the tanks will have an overflow rate of 240 and 470 gal/day/sq ft at average peak flows, respectively. Scum is to be removed from the surface of the final settling tanks by a rotary wiper arm. Scum from the surface of the settling tank is drawn over a short inclined beach and is discharged to a scum trough. The scum is then flushed to a scum well from which it is air lifted to the aerated sludge holding tanks. To maintain the activated sludge in an active condition, final sludge is removed from the settling tanks continuously. Sludge withdrawn from the final settling tanks is returned to the aeration tanks, at a rate to maintain a constant mixed liquor suspended solids and solids retention time in the aeration tanks, and to avoid excessive sludge depths in the settling tanks. Return sludge air lifts are used to return sludge to the head of the aeration tank. Excess sludge solids will be wasted from the settling tanks and air lifted to aerated sludge holding tanks to be concentrated prior to sludge dewatering. The concentrated sludge can be dried by mechanical means or pumped by portable pumps to a sludge drying bed. Sludge is allowed to dewater here by means of evaporation and drainage via an underground drainage system. The water, which drains from the sludge, is collected in a sump and is periodically pumped back to the influent side of the sewage treatment facility. Chlorination is used for disinfection of the final effluent at the new treatment facilities. Each treatment unit includes a separate chlorine contact zone of 170 cu ft which provides 15 min detention time and contact at the peak flow of 240,000 gal/day. Final effluent is dechlorinated to reduce total residual chlorine below permitted levels. Dechlorination chemical is added, based on total effluent flow rate, after the chlorine contact chamber and prior to the V-notch weir and effluent sample point. NMP Unit 1 UFSAR Section III III-30 Rev. 25, October 2017Each treatment unit contains an aerated sludge holding tank of approximately 2000 cu ft each. At double outage design flows, these tanks provide in excess of 30 days sludge storage. Each treatment unit is furnished with an aluminum geodesic dome cover for winterization protection. Each dome is equipped with two skylights and one gravity vent to provide natural lighting and ventilation. The walls of the treatment units are extended to support the domes and provide a workable clear headroom height along the interior circumference of the treatment unit. The domes are designed to be removable as a complete unit. The sludge drying bed structure is capable of being used as one drying bed or subdivided by removable dividers into multiple smaller drying beds. The drying bed structure is located just northwest of the sewage treatment facility. The structure is covered by a roof and has an 18-in curbing surrounding it. The curbing has multiple entrances to facilitate the removal of the dried sludge. 2.2.2 Ventilation System The STF is air conditioned and electrically heated. Unit air conditioners in the lab room only and heating coils for ventilation air are located throughout the facility where required. 2.2.3 Access Control The equipment house has no windows except in certain doors and a lock on the door prevents access by unauthorized personnel. 3.0 Energy Information Center The EIC is a single-story flat-roofed structure located on a slight promontory 1000 ft west and slightly south of the Station (Figure III-1). 3.1 Design Bases 3.1.1 Wind and Snow Loadings Exterior loadings for wind, snow, and ice used in design of the EIC meet all applicable codes as a minimum. The roof and its supporting structure are designed to withstand a loading of 40 psf of snow or ice. The walls and building structure are designed to withstand an external or internal loading of 40 psf NMP Unit 1 UFSAR Section III III-31 Rev. 25, October 2017of surface area, which is approximately equivalent to a wind velocity of 125 mph at the 30-ft level. 3.1.2 Pressure Relief Design There are no special pressure relief requirements for the EIC. 3.1.3 Seismic Design and Internal Loadings The EIC and components are designed as Class III structures with no special seismic criteria. The following design live loads were used in addition to the dead loads: Live load on stairways and all public areas except restrooms - 100 psf. Live load on all other floor areas including the classroom, offices and conference room - 60 psf. Allowable bearing pressure on undisturbed soil foundations of 1.5 tons/sq ft. Stresses in steel construction are those allowed by the AISC 1963 Specifications for the Design, Fabrication and Erection of Structural Steel for Buildings when using ASTM A36 Structural Steel. Stresses in concrete construction are those allowed by the ACI 318-63 Standard for 3000 psi concrete with intermediate grade new billet steel A-15. 3.1.4 Heating and Ventilation Heating and ventilation is provided for personnel comfort. 3.1.5 Shielding and Access Control No radioactivity is contained in or near the building; therefore, no shielding is required. 3.2 Structure Design 3.2.1 General Structural Features NMP Unit 1 UFSAR Section III III-32 Rev. 25, October 2017As shown on Figure III-1, the principal part of the building is in the form of a regular hexagon with sides 56-ft long. A wing of irregular shape but approximately 96-ft long by 36-ft and 45 1/2-ft wide extends to the west. The lobby occupies the full width of the southwest portion of the principal part of the building. To the rear of the lobby are a small theater, a room for a model of the Station and a room for various exhibits. The building's core, central to these rooms, contains a storage room, a projection room for the theater and stairs for access to the basement. Public restrooms and a women's lounge are located in the wing and adjoin the lobby on the left. The wing also contains a classroom, a conference room, offices, a central corridor, an extension of the main lobby and three secondary entrances to the building. The EIC building has a structural steel frame resting on a concrete substructure. Its exterior curtain walls are of concrete block with a veneer of native stone, trimmed with redwood, and well insulated. Interior walls are plastered metal or gypsum lath on steel studding. The roof is comprised of a bituminous waterproofing membrane on rigid insulation which is carried by metal roof decking and open web steel joist purlins, which are in turn supported by rolled steel girders and fascia beams. A concrete slab, hexagonally shaped in plan, about 30 ft in diameter and 4-in thick is centrally located on the roof to serve as a platform for the air conditioning condensers. 3.2.2 Heating and Ventilation System The EIC is air conditioned and electrically heated. Compressors, heat exchangers, heating coils for ventilation air and other mechanical equipment are located in equipment rooms in the basement. 3.2.3 Access Control Access to the EIC is from a separate road than that leading to the rest of the Station. Each room to which the public will be admitted has doors of ample width to the rooms adjoining on either side and, in addition, the theater and the model room each has its own exit door to the outside of the building. All these provide ample egress from any area for any conceivable emergency. NMP Unit 1 UFSAR Section III III-33 Rev. 25, October 2017 F. SCREENHOUSE, INTAKE AND DISCHARGE TUNNELS 1.0 Screenhouse The screenhouse adjoins the north wall of the reactor and turbine buildings and its superstructure is completely isolated from the reactor building. 1.1 Design Basis 1.1.1 Wind and Snow Loadings The wind and snow loadings for the screenhouse are the same as for the turbine building. 1.1.2 Pressure Relief Design There are no special pressure relief requirements for the screenhouse. 1.1.3 Seismic Design and Internal Loadings The screenhouse substructure has been designed to conform to the requirements for a Class I structure while loaded with any possible combination of filled and unwatered conditions of the channels located in this substructure. The superstructure is designed as a Class II structure as discussed on Page III-3 of the First Supplement to the PHSR. The seismic analysis resulted in the application of acceleration factors of 20.0 percent gravity horizontal and 10.0 percent gravity vertical. 1.1.4 Heating and Ventilation Heating and ventilation is provided for personnel comfort. 1.1.5 Shielding and Access Control No shielding is required. Normal access to the screenhouse is through the turbine building. 1.2 Structure Design The superstructure of the screenhouse is of framed structural steel supported on a reinforced concrete substructure which is founded on rock. The building has a flat roof consisting of cellular metal decking covered with insulation and a tar and NMP Unit 1 UFSAR Section III III-34 Rev. 25, October 2017felt roofing membrane. The two bays of the east wall, which are a continuation of an east wall of the turbine auxiliaries building extension, are of the same insulated sheet metal construction. The balance of the exterior wall, about 7/8 of the total, is of 8-in internally-insulated precast concrete panels corresponding with those in the base of the reactor building walls. Wall and roofing material and construction are identical with those used for the reactor and turbine buildings. The screenhouse substructure comprises channels for the flow of very large quantities of raw lake water, gates and stop logs for control of the flow, racks and screens for cleaning the water and pumps. The water channels are shown schematically on Figures III-19 and III-20. Five plain vertical gates near the north end of the substructure separate the channels from the tunnels. Gates A and B separate the intake tunnel from the forebay. Gate C separates the discharge channel from the discharge tunnel; gate E separates the discharge channel from the intake tunnel; and gate D separates the forebay from the discharge tunnel. Each of gates A, B, C, and D has a dedicated electric motor-driven hoist for raising, lowering, and maintaining position of the gates. Gate E is operated using a hydraulic ram system. Normal circulation is provided by opening gates A, B, and C with gates D and E closed. Reversed flow through the tunnels is obtained by closing gates A, B and C with gates D and E open. Tempering (partial recycle flow) is obtained by partially opening gate E with all other gates set for normal operation. The forebay and the secondary forebay are connected by three parallel cool water channels, in each of which are located trash racks, rack rakes and traveling screens to remove trash, water plants and fish from the water. Each of these channels has provisions for stop logs at each end so that any one of them may be segregated and unwatered for maintenance work without shutting down the Station. On the floor above the secondary forebay are mounted four containment spray raw water pumps and two emergency service water (ESW) pumps with a strainer for each. Also on this floor and above each of the three cool water channels are the screen wash pumps. Adjacent to the secondary forebay, on its south side and separated from it by channels fitted with stop log guides, are inlet chambers for the two circulating water pumps which provide water to the main condensers. By means of stop logs, either of these chambers can be isolated for unwatering and work on the corresponding pump. A lateral branch leads off NMP Unit 1 UFSAR Section III III-35 Rev. 25, October 2017to the east from the secondary forebay. Three chambers off this branch, separated from it by sluice gates, supply water to each of two service water pumps with strainers and a pair of fire pumps. One of these fire pumps is driven by an electric motor, the other by a diesel engine. The screenhouse is also equipped with a floor-operated electric overhead traveling bridge crane. This crane serves the various functions of placing and removing stop logs, and servicing the trash racks, rack rakes and traveling screens, maintenance of the two circulating water pumps and all pumps mounted above the secondary forebay. The service water pumps, their strainers, and the fire pumps are serviced for maintenance work by overhead beam runs, trolleys and hoists. 2.0 Intake and Discharge Tunnels As shown on Figure III-21, water is drawn from the bottom of Lake Ontario about two-tenths of a mile offshore and returned to the lake about one-tenth of a mile offshore. 2.1 Design Bases The water intake and discharge tunnels are designed to conform to the requirements for Class II structures. The intake and discharge tunnels are concrete-lined bores through solid rock. As such, they are highly rigid structures with extremely small natural periods of vibration and a seismic response of only 11 percent of gravity regardless of the damping factor. 2.2 Structure Design Water is admitted to the intake tunnel through a bellmouth-shaped inlet. The inlet is surmounted by a hexagonally-shaped guard structure of concrete, the top of which is about 6 ft above the lake bottom and 14 ft below the lowest anticipated lake level. The structure is covered by a roof of sheet piling supported on steel beams, and each of the six sides has a water inlet about 5-ft high by 10-ft wide, with the latter openings guarded by galvanized steel racks. This design provides for water to be drawn equally from all directions with a minimum of disturbance and with no vortex at the lake surface, and guards against the entrance of unmanageable flotsam to the circulating water system (CWS). The water drops through a vertical concrete-lined shaft to a concrete-lined tunnel in the rock, through which it flows to the NMP Unit 1 UFSAR Section III III-36 Rev. 25, October 2017foot of a concrete-lined vertical shaft under the forebay in the screenhouse. The foot of this shaft contains a sand trap to catch and store any lake-bottom sand which may wash over the sills of the inlet structure. The top of the shaft has a bell-mouthed discharge. Water is returned to the lake at a point about one-tenth of a mile offshore through a bell-mouthed outlet surmounted by a hexagonal-shaped discharge structure of concrete. The top of this structure is about 4 ft above lake bottom and 8 1/2 ft below the lowest anticipated lake level. The geometry of the structure closely resembles the inlet structure, although reduced in size. The six exit ports are about 3 ft high by 7 1/3 ft wide. The discharge tunnel from the screenhouse is identical in cross-section with the intake tunnel. The vertical shaft connecting the discharge tunnel with the discharge channel under the screenhouse also has a sand trap at its foot. Water is discharged directly to the vertical discharge shaft. A submerged diffuser in the vertical shaft ensures a good dilution before discharge to the lake. Samples are drawn at a lower point in the shaft. 3.0 Safety Analysis The selection and arrangement of equipment and components of the screenhouse and circulating water tunnels is based on the knowledge gained over many years of experience in the design, construction and operation of such facilities for coal-fired steam-electric stations. All components of the system which might possibly be subject to unscheduled outage, and by such outage affect the operability of the Station, are duplicated. In the case of the duplicate fire pumps, the prime movers are also totally independent. The gates are simple and rugged in construction, and their operation is simple and straightforward, with the possibility of inadvertent erroneous operation cut to a minimum. The pump suctions are amply submerged below the lowest low water surface elevation of the lake surface adjusted for the friction and velocity drops in the supply tunnel and channels. The supply of water by direct gravity from the lake is inexhaustible. The main portion of the superstructure, a single-story structure elastic frame of one bay width, has a relatively long natural period of vibration, and being bolted has a comparatively high damping factor. As a result, the dynamic loads which could be NMP Unit 1 UFSAR Section III III-37 Rev. 25, October 2017applied to it by wind pressure and also operation of the crane are more critical than those due to the seismic loading. Thus, while no dynamic analysis of the framing was required or made, it is quite probable that the building superstructure meets Class I conditions instead of only Class II, as specified in the First Supplement to the PHSR. Shearing forces in the walls and in the bottom chord plane of the roof truss system are resisted by systems of diagonal bracing. The sizes of the members of these systems were governed by detail and minimum allowable slenderness rather than by calculated forces, which resulted in excess strength being available in the system. G. STACK The stack is a freestanding reinforced-concrete chimney, 350-ft high, located 100 ft east of the northeast corner of the reactor building. 1.0 Design Bases 1.1 General The height of the stack and the velocity of discharge are to provide a high degree of dilution for routine or accidental Station effluents. This is discussed on Page IV-8 of the First Supplement to the PHSR. 1.2 Wind Loading Analysis shows that the loads due to seismic action are considerably greater than those which would be exerted by the velocity of wind for which the other Class I structures are designed: 125 mph at the 30-ft level. Since this is true for all levels of the stack (wind velocities and pressures varying according to elevation aboveground), lateral loads due to seismic forces govern the design. 1.3 Seismic Design The design and construction of the stack meet the seismic requirements of a Class I structure. Seismic forces applied are those obtained from the velocity and acceleration response spectra included in the First Supplement of the PHSR for a ground motion acceleration factor of 11 percent of gravity (Plate C-22). NMP Unit 1 UFSAR Section III III-38 Rev. 25, October 20171.4 Shielding and Access Control Shielding is required for the offgas and gland seal exhaust piping. Access is provided for inspection and maintenance during shutdown. 2.0 Structure Design The general features of the stack, including its principal dimensions, are shown on Figure III-22. It is a tapered monolithic reinforced-concrete tube resting on a massive concrete base which extends to sound rock. From this base it rises through the turbine auxiliaries building extension from which it is completely isolated structurally. The top of the stack is at el 611, or 212 ft 6 in above the top of the reactor building, the next highest structure in the Station. After filtration, all Station ventilation exhaust which is radioactively contaminated is brought to the stack through breaching, which is connected above the roof of the surrounding building. Two pipes, 6 in and 12 in in diameter, bring radioactively contaminated gases and vapors from the turbine shaft seals and from the condenser. These pipes enter the stack below the grade floor and turn up through encasing concrete to a terminal point at el 335, which is 20 ft above the top of the breaching entrance to the stack. At this point turbulence is high, which ensures best mixing and dilution of the contaminated gases. An "Isokinetic Probe" gas sampler is located within the stack with its orifices at el 535, or 76 ft below the top of the stack. This device is supported by a beam which spans the interior of the stack and cantilevers outside to facilitate withdrawal of the device for cleaning and maintenance. An opening is provided in the stack wall through which the device is installed. This opening is a 16-in diameter pipe sleeve with its outer end closed by a blind flange. A smaller adjoining opening makes it possible to measure the gas velocity profile in the stack or to visually inspect the probe without withdrawing it. The probe is connected to monitoring equipment located near the base of the stack by tubing which descends inside the stack. Access to the interior of the stack is through an airtight door from the basement of the surrounding building. Exterior access to the top of the stack and to four external platforms is from the roof of the building by means of a guarded ladder. At the probe level a small platform provides access and working area. NMP Unit 1 UFSAR Section III III-39 Rev. 25, October 2017Three other platforms completely surround the stack which provide access for external maintenance and painting of the stack. The stack is protected by four lightning rods and down conductors which are interconnected at the top, middle and bottom of the stack, then connected to the Station grounding grid. The structural reinforcing steel, platforms and ladder are in turn grounded by attachment to this system. The top of the stack is, in effect, an 8-ft 6-in inside diameter nozzle. For normal gas flows of 216,000 cfm, the corresponding velocity of the discharge jet is 63 fps. This relatively high velocity assures that the turbulence generated will thoroughly mix, dilute and disperse the discharged gas even at times of low wind velocity. 3.0 Safety Analysis 3.1 Radiology If during normal operation the stack were to be inoperative, there would be no serious radiological consequences for a period of time depending on the level of activity being released. If the stack were to remain inoperative for a significant length of time, the reactor would be shut down to prevent exceeding 10CFR20 limits. Exfiltration cases involving an inoperative stack are discussed in Section XV.

NMP Unit 1 UFSAR Section III III-41 Rev. 25, October 2017 If the stack fell within the northwest quadrant, the containment spray raw water, circulating water and service water pumps, as well as the lines from the diesel fire pumps, could be damaged. However, safe shutdown could still be afforded by use of the normal supplies of electric power and the emergency cooling system. H. SECURITY BUILDING WEST AND SECURITY BUILDING ANNEX The security building west and security building annex are located on the southwest corner of the Station security perimeter. See Figure III-1. Administrative offices are contained within these buildings for support of the duties associated with Station security. Because of the nature of this subject, a detailed description of these buildings will not be discussed in this document. For additional information regarding this subject, refer to the Station security plan. I. RADWASTE SOLIDIFICATION AND STORAGE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings Wind and snow loadings for the radwaste solidification and storage building (RSSB) are designed to meet or exceed those of the waste disposal building. 1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 1.3 Seismic Design and Internal Loadings(1) The foundation mat, structural walls, columns, floors and roof of the RSSB are classified as primary structural elements. All primary structural elements are seismically designed to withstand the effects of an operating basis earthquake (OBE) in accordance with Regulatory Guide (RG) 1.143. Secondary structure elements, including platforms, catwalks, pipe supports, equipment and vessel supports, and internal NMP Unit 1 UFSAR Section III III-42 Rev. 25, October 2017masonry walls, are classified as nonseismic-resistant items and are designed by conventional method. 1.4 Heating, Ventilation and Air Conditioning(2) The heating, ventilation and air conditioning (HVAC) and chilled water systems are designed for the following primary functional requirements: heat, ventilate and air condition the RSSB; remove airborne particulates from the RSSB atmosphere; prevent unfiltered exfiltration of airborne radioactivity from the building; prevent infiltration of airborne radioactivity into the RSSB control room and electrical room; control and provide a means for monitoring (via the main stack) the release of airborne radioactivity via the ventilation exhaust system; minimize the effects on the facility and its occupants from releases of radioactivity into the RSSB atmosphere; collect and filter air displaced via the vents from all RSSB tanks containing radioactive fluids; continuously purge the RSSB of truck exhaust fumes and other hazardous gases to ensure safe occupancy at all times. 1.5 Shielding and Access Control(3) Shielding is designed to limit radiation levels on the building exterior, in the control room, in the electrical room, stairwells, and the passageway to the truck bays. Access to the exterior of the RSSB is controlled by access to the protected area, which is controlled by Nuclear Security. Normal access to the building interior is via the waste building extension. Two exterior rollup doors allow access for vehicles to the two truck bays. Four exterior doors are normally locked and provide emergency egress. 2.0 Structure and Design Floor and roof plans and sections showing interior walls are shown on Figures III-3 through III-8. 2.1 General Structural Features(1) The RSSB is located to the east of, and is adjacent to, the existing offgas building, waste disposal building, and waste building extension of Unit 1. The arrangement of the RSSB can be considered as follows: process, handling and storage areas. NMP Unit 1 UFSAR Section III III-43 Rev. 25, October 2017This section is rectangular in shape and approximately 277 ft long below grade, 330 ft long above grade (north-south), and 61 ft wide (east-west). The majority of the primary structural components are reinforced concrete. The foundation mat is generally founded on top of bedrock. The finish grade and truck entrance and exit openings are at el 261'-0". The roof elevation is located at el 301'-2 1/2", with the material handling crane running longitudinally underneath the roof at el 292'-6 1/2". With the exception of a few feet around the perimeter, the crane can service the entire interior area of this section. Those portions of the RSSB which are classified as seismic-resistant elements are designed to maintain their structural integrity during and after all credible design loading phenomena, including OBE. Those items which are classified as seismic-resistant elements are the foundation base mat, structural concrete walls, floors and roof. Nonseismic-resistant structural elements are designed to maintain their structural function for all anticipated, credible design loading conditions encountered during construction, testing, operation, and maintenance of the facility. Those compartments containing large tanks (over 2,000 gal) of radioactive liquids are lined with steel to contain 1.5 tank volumes in the event of a tank rupture during a seismic event. During normal operation, maintenance, and loading and unloading operations, the structure provides sufficient environmental isolation to ensure that the exposure of plant operating personnel and the general public to radiation is ALARA. 2.2 Heating, Ventilation and Air Conditioning(2) Fresh air is filtered and conditioned and supplied to the control and electrical rooms, which are maintained at a slightly positive pressure with respect to other areas of the RSSB and the adjoining radwaste building. Air from other portions of the RSSB is not recirculated back to these areas. Air is recirculated within the RSSB and is processed through a filter system prior to reconditioning and redistribution. The recirculation filter system is comprised of the following primary filtration components: 1. Prefilters to remove larger particles to reduce dust loading on the high-efficiency particulate air (HEPA) filters. 2. HEPA filters with an individual efficiency of at least 99.97 percent. NMP Unit 1 UFSAR Section III III-44 Rev. 25, October 2017All RSSB ventilation exhaust air is processed through a filter train prior to discharging into the stack. The filter is comprised of the following primary filtration elements: 1. Prefilter to remove larger particles to reduce loading of the HEPA filters. 2. HEPA filters with an individual efficiency of at least 99.97 percent. 3. Two carbon adsorber sections for the removal of radioactive iodine from the exhaust stream. (Note: The charcoal adsorption capability provides added insurance that any release of iodine or other halogen activity will not result in offsite dose limits being exceeded. It is considered to be an enhancement, and not mandatory for the current use of the building for storage of waste and not for solidification of waste. Therefore, charcoal adsorption periodic testing is not required.) 4. Final HEPA filters with an individual efficiency of at least 99.97 percent. Air flow through the process areas of the RSSB is from areas of low radioactive contamination potential toward areas with increasingly higher contamination potential. Air from the two truck bays is ducted to the ventilation exhaust system rather than returned to the recirculating atmospheric cleanup system to prevent recirculation of truck exhaust fumes in the RSSB. The RSSB atmosphere is continuously purged (10,250 cfm) with clean outside air by operation of the fresh air supply and ventilation exhaust systems. Purge air from the process areas of the RSSB replaces the air drawn from the truck bays such that the entire building is purged via the exhaust from the truck bays. Radioactive tank vents are piped directly into the exhaust system upstream of the filter. Heating coils (electrical), cooling (chilled water), and fans are located downstream of the filter components to protect them from radioactive contamination. Supplemental heating is provided for the control and electrical rooms by duct heaters. Stair towers are provided with space heaters. Chilled water is produced in one of two 100-percent capacity water chillers and circulated by one of two 100-percent capacity chilled water pumps. Single failure of any one fan, heating coil or cooling coil may result in operating variations from the design basis; however, the overall effect with regard to the health and safety of the building occupants NMP Unit 1 UFSAR Section III III-45 Rev. 25, October 2017or the public will not be compromised. Fresh air inlet and ventilation exhaust penetrations through the RSSB outer walls are each fitted with two series mounted dampers designed to withstand a minimum of 3 psi pressure differential resulting from severe weather pressure conditions. All design and specification requirements are for nonseismic, nonnuclear safety-related systems and components. Instrumentation and control systems are provided to achieve required space temperature conditions and to maintain air flow requirements to provide acceptable building and process area pressure relationships. Relative humidity is not controlled, although it is maintained at reasonable levels by the HVAC system. All operating control functions are automatic. Temperature control systems in the fresh air supply and recirculating atmospheric cleanup systems are independent. Air flow control systems in the fresh air supply system and the exhaust ventilation system include interlock provisions to maintain pressure relationships upon de-energizing an exhaust or supply fan. Air flow controls of the recirculating atmospheric cleanup system are independent of the other systems. Redundant temperature sensing and control loops are provided in the fresh air supply and recirculating atmospheric cleanup system. Local instruments and remote indication and/or annunciation are provided. 2.3 Shielding and Access Control(3) The RSSB is designed to minimize exposure to plant personnel and the public by its location and design. The RSSB is located within the protected area and is heavily shielded by reinforced concrete. 3.0 Use The RSSB was constructed with the specific intent of providing onsite storage of low-level radioactive waste (LLW). The need to store LLW onsite is the result of the federal Low-Level Radioactive Waste Policy Act as amended in 1985, which initiated the process by which the three existing LLW disposal sites (Barnwell, SC; Beatty, NV; and Hanford, WA) would no longer be required to receive LLW. Although originally designed to store Unit 1 LLW, the RSSB is capable of providing interim storage of LLW produced at both Unit 1 and Unit 2. From a technical standpoint, the storage of Unit 2 waste at Unit 1 is considered acceptable based on the following: 1. The isotopic distributions of the waste stored in the RSSB from the two units are similar and expected to NMP Unit 1 UFSAR Section III III-46 Rev. 25, October 2017remain similar as both units have applied noble metals, inject depleted zinc, and inject low levels of hydrogen. 2. The selective storage of the high-activity LLW from both units in the RSSB (and the low-activity LLW at Unit 2) creates the potential for the storage of greater average activity concentration in the building, although not greater volume. However, since the RSSB was designed assuming the storage of incinerated resins which represent a bounding activity concentration, the building design is considered adequate for the combined storage from both units; 3. Total activity in the RSSB will ultimately be controlled per the Site radiation protection program to ensure that both onsite and offsite dose and dose rate limits are maintained; and 4. The transfer of by-product material between Unit 1 and Unit 2 will be conducted in accordance with approved radiation protection implementing procedures. Radioactive piping is routed through a shielded pipe tunnel and in shielded areas to limit exposure. Major pieces of equipment that can be significant sources of radiation exposure are each provided with a separate shielded cubicle. The storage vaults are shielded with 48 in of concrete in the storage zone (below crane). The roof is 24-in thick. The tank cubicles are shielded by 36 in of concrete. The east-west truck bay is equipped with a retracting shield door in the ceiling which mitigates albedo radiation in the truck bay from the storage vaults. The low-level storage room and the process equipment cubicle are equipped with sliding shield doors. Access is controlled administratively by the Unit 1 Radiation Protection Program. Physical control of high radiation areas is maintained in accordance with Technical Specifications. J. COLD STORAGE BUILDING The cold storage building (CSB) is located within the protected area on the east side of the station security perimeter. See Figure III-1. The building is designed to provide for the storage of radioactive materials outside of the main power block. Typical NMP Unit 1 UFSAR Section III III-47 Rev. 25, October 2017usage includes the storage of radioactive materials and equipment used to support station outages and receipt inspection of new fuel. The CSB is posted as a radiologically-controlled area (RCA). Access and storage of materials in this building are controlled by site Radiation Protection procedures. K. REFERENCES 1. Catalytic, Inc., Project No. 36700, System Description for Radwaste Solidification and Storage Building, Procedure No. 601 Revision 1, February 26, 1981. 2. Catalytic, Inc., Project No. 36700, System Description for Heating Ventilating and Air Conditioning (HVAC) and Chilled Water Systems, Procedure No. 204, 204.1 Revision 1, February 10, 1981. 3. Catalytic, Inc., Project No. 36700, System Description for Radiation Protection, Procedure No. 603 Revision 0, October 14, 1981. 4. NMPC Modification N1-98-016. Source Document: EY-OOSS FIGURE: I I I -1 Plot Plan NINE 1 NUCLEAR SCRIBAA%1.vs1s REPORT UPDATED SAFETY October 2015 UFSAR Rev. 24,

TURBINE BUILDING VENTILATION SYSTEM fEATlNG COIL ._ '--------ZONE *1 NOTE:ONEFANAECl'OFORNOAMALOPEAAT!ON mo FAN [5 EOLJIP'O 0/12SPEEOMOTOR,FANCN'ACITY 85.000/60,000 CFM (APPROX.I LARGE TURBINE BLDG. FL.EL.25'!1'-0' SMOKEZONE02 'l,il00CFMSLPPLY OFF-GAS BLJ!LOING TURB.BLDG. TU!1B.BL!X;. VENT SYS. EXHAUST FANS '*"'""""'"' $c-OO--B ._ __. TURB!NEBLOG * ... ________ SMOKEZONf."2 HI TEMP. SYS.TRIP .TLJl'CBINE:BLOG

• Fl.H.2:50'-0' SMOt>'.E20NE"3 TURB.BLQG,FL.EL.250'-0' SMOKEWNE*2-4500CFMSlF'PLY TURB.BLDG.FL..EL250'-0' SHOKE ZONE*\-1500 CFM SUPPLY OUTS10£FIXEO LOUVER &. SCAffN FIGURE Ill-12 UFSAR Rev1s1on 17 October-2001 LABORATORY AND RADIATION PROTECTION FACILITY VENTILATION SYSTEM I "' FROM MAIN TURBINE BLDG. VENTILATION SUPPLY CLOSE ON FAN TRIP !--------------.;; ;;;-;;;.;;;-----1 BOOSTER , .. HIGH EFFICIENCY FILTER 1,1111!1 CFM --------------------------------------------------------------, PIJTES. I. Tl£ ASSOCIATED l<<lV,TU: *TE ARE RETIRED IN PLACE. --, 0 *-""""""-* t BUILDINJ t ---E;v 0 HIGH LEVEL "' ___ ...,._--1@ EXHAUSTl()OO WITHHIGH EFFICIEl<<:YFILTER TO TL.llBINE BLDG. EXHAUST FAN I I Cl1JNTING ""'" FIGURE III-13 UFSAR REVISION 24 OCTOBER 212115 CONTROL ROOM VENTILATION SYSTEM -COllTIDL -II 111-*CY -___ 1 ___ ....... ;_. ___ _ !-----9 I II 9 ... Rlll'Al-I I I 11.u.u.v..:AI 1--' +----------+ PAI ....... *' .. ----I '---f----': f ** .. CHILL* *II .-s-. ...____tt l FIGURE 111-14 UFSAR REVISION 24 OCTOBER 2015 WASTE DISPOSAL BUILDING VENTILATION SYSTEM l)JlSIDE LO\JY(R AND SCREEN ------------' 1 CFUTURE) CEN1R!FUCiE El.26\'-il' WASTE COLLECTION . !RA!NTllNJ<5 5500 CFM---.. 300(0! ' ' ' : : '---"-" FILTER WASTE COLLECTOR FILTER WASTE OE:MINl':RALIZ£R LOW LEVEL *W>GE ..,,.,. WUTl..f!E> BLANKFLANClE FOR FUTURE OAUM flLllNG STllTJON ""' EFFJC!'ENCY "'ILTERS :---------------------------------------1-----------s t C: I :;J IBU!LOiNG ROOGHINt HIGH VENTILATION FILTER EFFJCIEHCY l'ILTER *u FILTER *n -BINDER MAIN &OVERFLOW COLLECTION CRIM l"Q..YM[R SOLlOlFICATION MIXER Ct<TRO.WAS1'£ METERlMCTAI<< S'YS.STAATPERMISS!VE ---------------., ' EICHM.61 FAN Ga9 CF" WASTE DISPOSAi.. BLDG.fl.*El..26t'-9' FIGURE III-15 UFSAR REVISION 17 OCTOBER 2001 WASTE DISPOSAL BUILDING EXTENSION VENTILATION SYSTEM I : I : : Ct:NCENTRATEO llAS!Ellltt<;"12 '"'" DECON. ""' COll:EHTRATEO WASTEPUMP*l2 '"'" EL.26t'*r se0 CFM Y(lolTFl!OM llASfE OISPOSPL SYSTEM AIR80R!£ACTJVITYl"IOHITOA I 1 I I 1 I .--------.t $ b ,t-----t : I I I : I $ b i_ £>>WJ'li1 FAN "14 '"'°"""' FIGURE III-16 UFSAR REVISION 18 OCTOBER 2003 OFF GAS BUILDING VENTILATION SYSTEM PRElllOSOllBER ORAIMAREA FL.ODii. AREll ELECTJl!CeDlt.L'll *12 llREA ..,,,,. ""'""" ,,, .. (l.Z61'*r 29' CFH I :---------------------------I I I < I I ! I ! ! i: I :---------------------------I I I I I I [ f __ .... ,,,. FIGURE III-17 UFSAR Revision 16 November 1999 TECHNICAL SUPPORT CENTER VENTILATION SYSTEM NOTES: 10021 CFM LOLIYER EHERCiENCY HOOE !il00 CfM OUTSIDE LOVVER EMERGENCY MOl:e Cll MECHANICAL LINKAGE DISCONNECTED AND DAMPER LOCKED OPEN AT 3000 CFM t 10%. AUTOMATIC BALANClNG CAPABILITY DEFEATED. PA£*FILTER l-£PA*FILT£R CHARCOAL FILTEl:I dv HEPA*F!LTER NORMALLY OPEN CLOSES ON EMERGENCY MOOE SHOKE DETECTOR COIL 7.5 KW FILTER OX COOLING COIL 7'!1!5 CFM DAMPER CFM NORMAL 41!S CFM EMERGENCY !Sflfl CFM EXHAUST MIN. 'lee CFM EXHAUST HAI. --RETUIULLY CPrn CLOSES OH EMERGENCY 1-00E UfSAR Rev. 14 (June 11N)

DISCHARGE TUNNEL CIRCULATING WATER CHANNELS UNDER SCREEN AND PUMP HOUSE NORMAL OPERATION t INTAKE TUNNEL ,--.:-1 I I I I I I PRIMARY FORE BAY SECONDARY FOREBAY NORMAL OPERATION: GATES A, B & C *OPEN GATES D & E *CLOSED FIG\JRE 11-19 UfSAR Rev. 14 &June 1 IM)

i " *J* j CIRCULATING WATER CHANNELS UNDER SCREEN AND PUMP HOUSE SPECIAL OPERATIONS REVERSED FLOW IN INTAKE AND -1 DISCHARGEI"" I TUNNELS I I ! I I L.. -J I I I 1 I I LI FROM AND TO CONDENSERS . ,..-ri I I I I I I I I ,.-t--i I I I I I I PART PLAN AT UPPER LEVEL REVERSED FLOW OPERATION: GATES, A, B & C CLOSED GATES D & E OPEN I I ONE CIRCULATING WATER PUMP : WELL & ONE SERVICE WATER ' FIGURE 11-20 UFSAR Rev. 14 (June 1996)

INTAKE AND DISCHARGE TUNNELS PLAN AND PROFILE ';'; \ 1195'-0" PROFILE ALONG INTAKE TUNNEL PROl=ILE. ALOf.JG DISCHARGE. Tl:JNNE.L 0 QQAPHIC l= 200 j?O E.LE. VA TION tt-JTAKe I\;\) \; \; \) ', ( SE.CTION 1* 1 TYPIC"-L ---------0 5 10 GRAPl-11C 0 z 4-.. 0 <> GQAPHIC SECTIOl-J 2*2 TYPICAL ---------------, I 1, __ j; PLAt-J E.LCVA.TION DISCHARGE. FIGURE IIl-21 UFSAR Rev. 17 (October 2001) Source Documents: C-15448-C C-15449-C C-15450-C ST ACK -PLAN AND ELEVATION u*

• F '° ao nEvATl()l.I GRAPHIC PLAN Pl.ATl"ORM C0Nc:.. ENvtl.OP[. tL 372 '-0" EL.1SO'O"Tl>tL$0' .R2.97'Cf _...tbJM::O" 110Ck PRO&t tL.221:"" ..J.... ................ ....,...., ........ i @ LOOKING FIGURE 111-22 UFSAR Rev. 14 (June 1996)

NMP UNIT 1 UFSAR Section IV EF IV-1 Rev. 25, October 2017 LIST OF EFFECTIVE FIGURES SECTION IV Figure Number Revision Number IV-1 14 IV-2 14 IV-3 14 IV-4 15 IV-4a 15 IV-5 14 IV-6 14 IV-7 16 IV-7a 16 IV-8 14 IV-9 14 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section IV IV-i Rev. 25, October 2017 SECTION IV REACTOR A. DESIGN BASES 1.0 General 2.0 Performance Objectives 3.0 Design Limits and Targets B. REACTOR DESIGN 1.0 General 2.0 Nuclear Design Technique 2.1 Reference Loading Pattern 2.2 Final Loading Pattern 2.2.1 Acceptable Deviation From Reference Loading Pattern 2.2.2 Reexamination of Licensing Basis 2.3 Refueling Cycle Reactivity Balance 3.0 Thermal and Hydraulic Characteristics 3.1 Thermal and Hydraulic Design 3.1.1 Recirculation Flow Control 3.1.2 Core Thermal Limits 3.1.2.1 Excessive Clad Temperature 3.1.2.2 Cladding Strain 3.1.2.3 Coolant Flow 3.2 Thermal and Hydraulic Analyses 3.2.1 Hydraulic Analysis 3.2.2 Thermal Analysis 3.2.2.1 Fuel Cladding Integrity Safety Limit Analysis 3.2.2.2 MCPR Operating Limit Analysis 3.3 Reactor Transients 4.0 Stability Analysis 4.1 Design Bases 4.2 Stability Analysis Method 5.0 Mechanical Design and Evaluation 5.1 Fuel Mechanical Design 5.1.1 Design Bases 5.1.2 Fuel Rods 5.1.3 Water Rods 5.1.4 Fuel Assemblies 5.1.5 Mechanical Design Limits and Stress Analysis 5.1.6 Relationship Between Fuel Design Limits and Fuel Damage Limits NMP Unit 1 UFSAR Section Title Section IV IV-ii Rev. 25, October 2017 5.1.7 Surveillance and Testing 6.0 Control Rod Mechanical Design and Evaluation 6.1 Design 6.1.1 Control Rods and Drives 6.1.2 Standby Liquid Poison System 6.2 Control System Evaluation 6.2.1 Rod Withdrawal Errors Evaluation 6.2.2 Overall Control System Evaluation 6.3 Limiting Conditions for Operation and Surveillance 6.4 Control Rod Lifetime 7.0 Reactor Vessel Internal Structure 7.1 Design Bases 7.1.1 Core Shroud 7.1.2 Core Support 7.1.3 Top Grid 7.1.4 Control Rod Guide Tubes 7.1.5 Feedwater Sparger 7.1.6 Core Spray Spargers 7.1.7 Liquid Poison Sparger 7.1.8 Steam Separator and Dryer 7.1.9 Core Shroud Stabilizers 7.1.10 Core Shroud Vertical Weld Repair 7.2 Design Evaluation 7.3 Surveillance and Testing C. REFERENCES NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section IV IV-iii Rev. 25, October 2017 IV-1 LIMITING POWER/FLOW LINE (TYPICAL) IV-2 thru FIGURES DELETED IV-3 IV-4 TYPICAL CONTROL ROD - ISOMETRIC IV-5 FIGURE DELETED IV-6 CONTROL ROD DRIVE AND HYDRAULIC SYSTEM IV-7 CONTROL ROD DRIVE ASSEMBLY IV-8 TYPICAL CONTROL ROD TO DRIVE COUPLING - ISOMETRIC IV-9 REACTOR VESSEL ISOMETRIC NMP Unit 1 UFSAR Section IV IV-1 Rev. 25, October 2017 SECTION IV REACTOR A. DESIGN BASES 1.0 General The functional requirements for the reactor core are as follows: 1. The core is designed for steady-state operation at 1850 MWt. 2. The equilibrium reload batch average discharge exposure is projected at approximately 44,000 MWD/ST. 3. The reactor is currently operating on a 24-month refueling cycle. Approximately 33 percent of the core is changed out each refueling. 2.0 Performance Objectives Those performance objectives pertinent to the safe operation of the reactor are as follows: 1. Seismic criteria as described in Section III. 2. Fuel thermal limits (i.e., minimum critical power ratio (MCPR), maximum linear heat generation rate (MLHGR), and maximum average planar linear heat generation rate (MAPLHGR)) preclude excessive fuel rod failures and maintain operations within 10CFR50.46 limits. 3. The fuel rod cladding is designed to contain fission gas release. 4. The reactor is designed so that there will be no inherent tendency for undamped power oscillations. 5. Power excursions from a reactivity addition accident will not result in excessive fuel rod failures, rupture the pressure vessel or impair operation of required safeguards equipment. NMP Unit 1 UFSAR Section IV IV-2 Rev. 25, October 2017 6. Reactivity shutdown capability is provided to make and hold the core adequately subcritical by control rod action, at the conditions of peak core reactivity, assuming that any one control rod is fully withdrawn and unavailable for use. 7. Redundant backup reactivity shutdown capability is provided independent of normal control. This additional reactivity control has the capability to shut down the reactor from any operating condition. 3.0 Design Limits and Targets The design limits and targets for the core are summarized in the Supplemental Reload Licensing Report (SRLR)(1) and in the Core Operating Limits Report (COLR). Each fuel assembly and its components are designed to withstand: 1. The predicted thermal, pressure, and mechanical interaction loadings occurring during startup testing, normal operation, and abnormal operational transients without impairment of operational capability; 2. Loading predicted to occur during handling without impairment of operational capability; 3. In-core loading predicted to occur from an operating basis earthquake (OBE), occurring during normal operating conditions, without impairment of operational capability. B. REACTOR DESIGN 1.0 General The nuclear performance characteristics of the core provide a dynamic response which: 1. Has a strong negative reactivity feedback under severe reactivity addition transients. 2. Contributes negative reactivity feedback consistent with the requirements of overall plant nuclear-hydrodynamic stability. 3. Has a reactivity response which regulates or damps changes in power level and spatial distribution of NMP Unit 1 UFSAR Section IV IV-3 Rev. 25, October 2017power production in the core to a level consistent with safe and efficient operation. Having these characteristics at all operating conditions of the core and at all states of fuel exposure contributes to satisfying performance objectives "4" and "5." Specifically, characteristic "1" is a major factor through the Doppler and moderator coefficients in providing shutdown mechanisms in the event of a reactivity excursion. Characteristics "2" and "3" are obtained to assure, along with other parameters, that there will be no inherent tendency for undamped oscillations. The dynamic behavior of the core is characterized in terms of several reactivity coefficients. These are: 1. Fuel temperature or Doppler coefficient. 2. Steam void coefficient. 3. Overall moderator temperature coefficient. The simultaneous effect of all three coefficients is termed the power coefficient. In UO2 fuel the Doppler coefficient provides potential for a large instantaneous negative reactivity feedback to any power rise, either gross or locally, of the core. The magnitude of the Doppler coefficient is inherent in the fuel design and does not vary significantly among light-water moderated UO2 fuel designs having low enrichment. Since a number of reactor parameters, including the void coefficient, contribute to stability requirements, no specific void coefficient value can be used as a design basis. However, to assure stability, the void coefficient during power operation must not become too negative. The water-to-fuel ratio provides a void coefficient consistent with other parameters in meeting stability requirements. A second requirement is imposed on the void coefficient, or more precisely to the moderator density coefficient inside the fuel assemblies. In Doppler-terminated or controlled transients, this moderator coefficient is relatively slow acting due to the long heat transfer time constant of the fuel. Nevertheless, it should not result in a significant positive reactivity contribution to the core as heat transfers from fuel to coolant. NMP Unit 1 UFSAR Section IV IV-4 Rev. 25, October 2017The overall moderator temperature coefficient includes temperature effects of the moderator in the gaps between flow channels. This coefficient is even slower acting than the void coefficient as it takes on the order of minutes for the water gaps to reach temperature equilibrium with the circulating coolant. Because of the relatively slow response of the water gap temperature during transients, the reactivity feedback due to density changes in the gaps is negligible. Due to design requirements on the magnitude and sign of the void coefficient, the temperature coefficient inherently becomes and remains negative before normal reactor operating temperature is reached. Finally, perturbations of reactor power level result in shifts of power distribution in the core due to the effects of xenon variations. The inherent nuclear characteristics of the core lead to strong damping of such oscillations. Such damping is provided by the operating power coefficient of the core. Operating experience and analysis has indicated that in large boiling water reactors (BWRs), a power coefficient more negative than about -0.01 k/k/P/P provides damping of xenon-induced power shifts to the point that they can be maintained within normal operating limits by minor control rod adjustment.(2) 2.0 Nuclear Design Technique The nuclear evaluation of reload cores is performed in accordance with Reference 3. Details of the analyses are also contained in Reference 3. Most of the lattice analysis is performed during the fuel bundle design process. The results of these single bundle calculations are reduced to libraries of lattice reactivities, relative rod powers, and few group cross sections as functions of instantaneous void, exposure, exposure-void history, control state, and fuel and moderator temperatures. These libraries of information are then used as input to the full-core analysis. The primary lattice characteristics of concern are: 1. Cold clean K data 2. Hot K as a function of exposure and void fraction 3. Local peaking factor 4. Doppler coefficient NMP Unit 1 UFSAR Section IV IV-5 Rev. 25, October 2017The core analysis is unique for each reload. This is performed in the months preceding the reload to demonstrate that the core meets all applicable safety limits. The principal tool used in the core analysis is the Three-Dimensional Boiling Water Reactor Simulator Code,(4) which computes power distributions, exposure, and reactor thermal-hydraulic characteristics with spatially varying voids, control rods, burnable poisons and other variables. Since the core analysis for an upcoming reload is performed prior to the end of the current cycle, it must be based on assumptions about the condition of the core that is expected. The analysis develops a reference loading pattern based on the best possible prediction of the core condition and on the desired reload core energy, and checks that against applicable safety limits. When the actual reload is completed, the "as-loaded" core may not be identical to the reference loading pattern. To assure safety limits are not violated by the as-loaded configuration, key parameters of the two patterns are compared. Conservative bounds on the amount and restrictions on the manner by which these key parameters may vary from the reference are imposed on the as-loaded core. When a reload core cannot meet these restrictions, all the affected licensing parameters must be reexamined to assure that there is no adverse impact; only when this reexamination has been completed, and it has been established that the as-loaded core satisfies the licensing basis, will the core be operated. 2.1 Reference Loading Pattern The reference loading pattern is analyzed under the steady-state condition to determine shutdown margin (including the liquid poison system) and core average reactivity coefficients. The following parameters are computed for the reference loading pattern and are included in the reload analysis process. 1. Core effective multiplication 2. Control system worth 3. Reactor shutdown margin 4. Shutdown capability of the liquid poison system 5. Moderator void coefficient 6. Doppler coefficient NMP Unit 1 UFSAR Section IV IV-6 Rev. 25, October 2017 7. Moderator temperature coefficient 2.2 Final Loading Pattern 2.2.1 Acceptable Deviation From Reference Loading Pattern Sensitivity studies have been conducted which determined how and by how much the key parameters may be allowed to vary without adversely affecting the licensing analysis. The following parameters are routinely checked for every reload: 1. Core average end-of-cycle (EOC) exposure 2. Core average EOC axial exposure distribution 3. Number of reload bundles 4. Type and number of exposed bundles 5. Locations of reload bundles 6. Locations of exposed bundles 7. Exposure and location history of edge bundles 8. Symmetry 9. Shutdown margin If the deviations between the reference loading pattern and the final loading pattern are found to be acceptable, the core may be operated. If the deviations are unacceptable, then a reexamination of the final loading pattern is required. 2.2.2 Reexamination of Licensing Basis The parameters studied in the reexamination of the final loading pattern are as follows: 1. SCRAM reactivity insertion 2. Dynamic void coefficient 3. Peak fuel enthalpy during rod drop accident 4. Cold shutdown margin NMP Unit 1 UFSAR Section IV IV-7 Rev. 25, October 2017 5. Change in critical power ratio (CPR) due to a misloaded fuel assembly 6. Rod block monitor response to a rod withdrawal error (RWE) 2.3 Refueling Cycle Reactivity Balance For each fuel cycle, the reactivity balance at the beginning of cycle (BOC) and the R-value are calculated. The results are as follows: Cold Clean Core, 20°C Keff, Uncontrolled SRLR(1) Keff, Fully Controlled SRLR(1) Keff, Strongest Control Rod Out SRLR(1) R, Maximum Increase in Cold Core SRLR(1) Reactivity with Exposure, Keff The R-value is the difference in reactivity with the strongest control rod out at BOC, and the reactivity with the maximum calculated strongest rod out at any exposure. 3.0 Thermal and Hydraulic Characteristics 3.1 Thermal and Hydraulic Design 3.1.1 Recirculation Flow Control Reactor power can be controlled over an approximate 50-percent power range, but no lower than about 40 percent of full power, by adjustment of the reactor recirculation flow with no control rod movement. Reactor power change is accomplished by utilizing the large negative power coefficient characteristic of BWR designs. To increase reactor power, recirculation flow is increased which reduces the void accumulation in the core by removing the steam at a faster rate. A positive reactivity input is balanced by negative reactivity effects of higher fuel temperature and new void formation. When these effects balance out, the reactor will be operating at a higher power level. The feedwater level controller increases the feedwater flow to match the increased steam generation. Conversely, when a power NMP Unit 1 UFSAR Section IV IV-8 Rev. 25, October 2017reduction is required, recirculation flow is reduced. A typical relationship between coolant flow rate and reactor power is shown on Figure IV-1. The reactor recirculation pump speed control is discussed in detail in Section VIII-B. 3.1.2 Core Thermal Limits Core thermal limits are based on two potential thermal damage modes: excessive cladding temperature and excessive cladding strain. Fuel damage is defined as a loss of cladding integrity allowing release of fission products to the coolant. The clad temperature failure mode is dependent on MCPR; the clad strain mode is dependent on the magnitude of the linear heat generation rate (LHGR). Clad failure due to high temperature following a loss-of-coolant accident (LOCA) is dependent on MAPLHGR. 3.1.2.1 Excessive Clad Temperature During normal operation and operational transients, nucleate boiling occurs in the core. Nucleate boiling is characterized by a small clad-to-coolant temperature drop so that resultant clad temperatures are only slightly above the coolant temperature. At sufficiently higher power levels, the transition boiling mode would be initiated. Transition boiling is accompanied by cladding temperature fluctuations. The bundle power, at which some point within the assembly experiences onset of transition boiling, is termed the critical power. If power is increased sufficiently beyond this point, the film boiling mode would occur and result in potential clad perforations. A figure of merit utilized for establishing reactor operating limits is the CPR. This is the ratio of the critical power to the operating bundle power. The critical power is determined at the same mass flux, inlet temperature, and pressure which exists at the specified reactor condition. Thermal margin is stated in terms of the MCPR which corresponds to the most limiting fuel assembly in the core. To ensure that adequate margin is maintained, the following transient design requirement was chosen: Moderate frequency transients caused by a single Operator error or equipment malfunction shall be limited such that, considering uncertainties in manufacturing and monitoring the core operating state, more than 99.9 percent of the fuel rods would be expected to avoid boiling transition. NMP Unit 1 UFSAR Section IV IV-9 Rev. 25, October 2017 Using this basic design requirement, both normal operating and transient thermal limits in terms of MCPR are derived. These limits are determined in accordance with the methods described in Reference 3. With each reload, compliance with the limits on MCPR is verified by transient analyses. For each cycle, the most limiting transient MCPR is reported in the SRLR(1) for that cycle and updated in the COLR. Clad failure due to high temperature following a postulated LOCA is a function of average heat generation rate of all rods of a fuel assembly. Average planar linear heat generation rate (APLHGR) is the parameter which describes the potential for that failure and a limit on MAPLHGR is established for core operation. The limits on APLHGR as a function of exposure are shown in the SRLR(1) and in the COLR. These limits assure that 10CFR50 Appendix K temperature limits are not exceeded. 3.1.2.2 Cladding Strain Rupture of fuel cladding due to excessive cladding strain may result from the combined effects of cladding creep-down, fuel pellet irradiation swelling, and fuel pellet-fragment outward relocation. A value of 1% plastic strain of Zircaloy cladding is conservatively defined in Reference 20 as the limit below which fuel damage from overstraining the fuel cladding is not expected to occur. A cladding strain evaluation is performed using PRIME thermal-mechanical performance model in conjunction with worst tolerance assumptions as described by Reference 3. That evaluation defines a Mechamical Overpower (MOP) limit that ensures the circumferential strain will not exceed the specified strain limit for the maximum duty rod. Each cycle, fuel rods are evaluated for conformance to the MOP limit as reported in the SRLR, and this ensures that fuel rod failure due to pellet-clad mechanical interaction will not occur. 3.1.2.3 Coolant Flow Coolant for the core flows from the discharge of the pumps through the bottom plenum region of the reactor to the fuel assembly inlet orifices at the bottom of the core. Orificing of NMP Unit 1 UFSAR Section IV IV-10 Rev. 25, October 2017the core inlet is employed to achieve a more uniform exit quality distribution from the core and to provide additional flow in the high-powered assemblies. All fuel positions are orificed with the more restrictive orificing in the periphery of the core. In general, the increase in single-phase DP improves the stability and makes channel flow less sensitive to channel power. Except for a small fraction which bypasses the fuel assemblies and cools the core components between the fuel channels, the recirculation flow travels vertically upward within the fuel assemblies to cool the fuel rods. 3.2 Thermal and Hydraulic Analyses 3.2.1 Hydraulic Analysis Core steady-state thermal-hydraulic analyses are performed using a model of the reactor core, which includes hydraulic descriptions of orifices, lower tie-plates, fuel rods, fuel rod spacers, upper tie-plates, the fuel channel, and core bypass flow paths. The orifice, lower tie-plate, fuel rod spacers, upper tie-plate, and, where applicable, holes in the lower tie-plate, are hydraulically represented as being separate, distinct local losses of zero thickness. The fuel channel cross section is represented by a square section with enclosed area equal to the unrodded cross-sectional area of the actual fuel channel. The flow distribution to the fuel assemblies and bypass flow paths is calculated on the assumption that the pressure drop across all fuel assemblies and bypass flow paths is the same. This assumption has been confirmed by measuring the flow distribution in BWRs. The components of bundle pressure drop considered are friction, local, elevation, and acceleration. Pressure drop measurements made in operating reactors confirm that the total measured core pressure drop and calculated core pressure drop are in good agreement. There is reasonable assurance, therefore, that the calculated flow distribution throughout the core is in close agreement with the actual flow distribution of an operating reactor. An iteration is performed on flow through each flow path (fuel assemblies and bypass paths), which equates the total differential pressure (plenum to plenum) across each path and matches the sum of the flows through each path to the total core flow. The total core flow less the control rod cooling flow enters the lower plenum. A fraction of this passes through various bypass paths. The remainder passes through the orifice in the fuel support (experiencing a pressure loss), where more NMP Unit 1 UFSAR Section IV IV-11 Rev. 25, October 2017flow is lost through the fit-up between the fuel support and the lower tie-plate, and through the lower tie-plate holes into the bypass region. The majority of the flow continues through the lower tie-plate (experiencing a pressure loss) where some flow is lost through the flow path defined by the fuel channel and lower tie-plate into the bypass region. This bypass flow is restricted on those fuel assemblies with finger springs. Within the fuel assembly, heat balances on the active coolant are performed nodally. Fluid properties are expressed as the bundle average at the particular node of interest and are based on 1967 International Standard Steam-Water Properties. In evaluating fluid properties, a constant pressure model is used. The relative radial and axial power distributions are used with the bundle flow to determine the axial coolant property distribution, which gives sufficient information to calculate the pressure drop components within each fuel assembly type. When the equal pressure drop criterion described above is satisfied, the flow distributions are established. When flow distributions have been determined, limits on plant operation are established to assure that the plant can be safely operated and not pose any undue risk to the health and safety of the public. This is accomplished by demonstrating that radioactive release from plants for normal operation, abnormal operational transients, and postulated accidents meet applicable regulations in which conservative limits are documented. This conservatism is augmented by using conservative evaluation models and observing limits which are more restrictive than those documented in the applicable regulations. 3.2.2 Thermal Analysis The objective for normal operation and transient events is to maintain nucleate boiling and, thus, avoid a transition to film boiling. Operating limits are specified to maintain adequate margin to the onset of the boiling transition. Both the transient (safety) and normal operating thermal limits in terms of MCPR are derived from this basis. 3.2.2.1 Fuel Cladding Integrity Safety Limit Analysis The generation of the MCPR limit requires a statistical analysis of the core near the limiting MCPR condition. The statistical analysis is used to determine the MCPR corresponding to the transient design requirement of paragraph 3.1.2.1. This MCPR NMP Unit 1 UFSAR Section IV IV-12 Rev. 25, October 2017established fuel cladding integrity safety limit applies not only for core-wide transients, but is also conservatively applied to the localized RWE transient. The statistical analysis utilizes a model of the BWR core which simulates the core monitoring function. This code produces a CPR map of the core based on inputs of power distribution and flow and on heat balance information. Details of the procedure are documented in Reference 17. Power distribution uncertainties used in the cycle-specific statistical analysis are presented in Reference 18. The minimum allowable CPR is set to correspond to the criterion that 99.9 percent of the rods are expected to avoid boiling transition by averaging the means of the distributions formed by all the trials. Cycle-specific analyses have been performed, as described in Reference 3, which provide conservative safety limit MCPRs. The results of the analysis are summarized in Reference 1. The results of the analyses show that at least 99.9 percent of the fuel rods in the core are expected to avoid boiling transition if the MCPR is greater than or equal to the fuel-specific safety limit, as specified in the SRLR (Reference 1) and the COLR. The fuel cladding integrity safety limit MCPR is contained in the COLR. 3.2.2.2 MCPR Operating Limit Analysis A MCPR operating limit is established to ensure that the Fuel Cladding Integrity Safety Limit is not exceeded for any moderate frequency transient. This operating requirement is obtained by addition of the absolute, maximum CPR value for the most limiting transient (including any imposed adjustment factors), from rated conditions postulated to occur at the plant to the fuel cladding integrity safety limit. There are eight nuclear system parameter variations or transients which could pose potential deleterious effects to the nuclear steam supply system (NSSS). These parameter variations are: 1. Nuclear system pressure increase - threatens to rupture the reactor coolant pressure boundary (RCPB) from internal pressure. Also, a pressure increase NMP Unit 1 UFSAR Section IV IV-13 Rev. 25, October 2017collapses the voids in the moderator. This causes an insertion of positive reactivity which may result in exceeding the fuel cladding safety limits. 2. Reactor vessel water (moderator) temperature decreases - results in an insertion of positive reactivity as density increases. Positive reactivity insertions threaten the fuel cladding safety limits because of high power. 3. Positive reactivity insertion - is possible from causes other than nuclear system pressure or moderator temperature changes. Such reactivity insertions threaten the fuel cladding safety limits because of higher power. 4. Reactor vessel coolant inventory decrease - threatens the fuel as the coolant becomes unable to maintain nucleate boiling. 5. Reactor core coolant flow decrease - threatens the fuel cladding safety limits as the coolant becomes unable to maintain nucleate boiling. 6.Reactor core coolant flow increase - reduces the void content of the moderator, resulting in a positive reactivity insertion. The resulting high power may exceed fuel cladding safety limits. 7. Core coolant temperature increase - could exceed fuel cladding safety limits. 8. Excess of coolant inventory - could result in damage resulting from excessive carry-over. Of these parameter variations, only a few are characteristic of operating transients which would result in a significant reduction in MCPR. To determine the limiting transient events, the relative dependency of CPR upon various thermal-hydraulic parameters was examined. A sensitivity study was performed to determine the effect of changes in bundle power, bundle flow, subcooling, R-factor, and pressure on CPR for the 8x8 fuel design. Results of the study indicate that CPR is most responsive to fluctuations in the R-factor and bundle power. A slight NMP Unit 1 UFSAR Section IV IV-14 Rev. 25, October 2017sensitivity to pressure and flow changes and relative independence to changes in inlet subcooling was also shown. The R-factor is a function of bundle geometry and local power distribution and is assumed to be constant throughout a transient. Therefore, transients which would be limiting because of MCPR would primarily involve significant changes in power. Based on this, the transients most likely to limit operation because of MCPR considerations are: 1. Turbine trip without bypass, or generator load rejection without bypass. 2. Loss of feedwater heating, or inadvertent high-pressure coolant injection (HPCI) startup. 3. Feedwater controller failure (maximum demand). 4. Control RWE. 5. Recirculation flow controller failure - increasing flow and an inadvertent startup of a cold recirculation loop as related to Kf curve. The above transients are reevaluated for each reload core. The results of the analysis are summarized in the SRLR(1) and are used to establish the most limiting transient and the MCPR operating limit. 3.3 Reactor Transients Core-wide rapid pressurization events (turbine trip without bypass and feedwater controller failure) are analyzed using the system model documented in Reference 5. The ODYN code contains a one-dimensional representation of the reactor core which is coupled to the recirculation and control system model. The integrated model is based on one-dimensional reactor kinetics, multinoded thermal-hydraulic and heat transfer relationships, and mechanical kinetic equations of the equipment. ODYN contains a refined reactor core description and a detailed steam line model to simulate pressure dynamics during a transient. For the slower core-wide transients, loss of feedwater heating is analyzed using either the steady-state 3-D BWR Simulator Code(4), or the REDY Transient Model(6). A more thorough description of the transients analyzed is given in Section XV-B.3.0. 4.0 Stability Analysis NMP Unit 1 UFSAR Section IV IV-15 Rev. 25, October 2017 4.1 Design Bases Three types of stability are considered in the design of BWRs: 1) reactor core (reactivity) stability; 2) channel hydrodynamic stability, and 3) total system stability. A stable system is analytically demonstrated if no inherent limit cycle or divergent oscillation develops within the system as a result of calculated step disturbances of any critical variable, such as steam flow, pressure, neutron flux, or recirculation flow. The criteria for evaluating reactor dynamic performance and stability are stated in terms of two compatible parameters. First is the decay ratio, x2/x0, which is the ratio of the magnitude of the second overshoot to the first overshoot resulting from a step perturbation. A plot of the decay ratio is a graphic representation of the physical responsiveness of the system which is readily evaluated in a time-domain analysis. Second is the damping coefficient, n, the definition of which corresponds to the dominant pole pair closest to the imaginary axis in the s-plane for the system closed-loop transfer function. As n decreases, the closed-loop roots approach the imaginary axis and the response becomes increasingly oscillatory. 4.2 Stability Analysis Method The stability analysis methods for the Nine Mile Point Nuclear Power Station are documented in GENE-A13-00360-02, "Application of Stability Long-Term Solution Option II to Nine Mile Point Nuclear Station Unit 1," August 30, 1995; NEDC-32992P-A, "ODYSY Application for Stability Licensing Calculations," July 2001; and NEDE-33213P-A, "ODYSY Application for Stability Licensing Calculations Including Options I-D and II Long-Term Solutions," April 2009. GE6/7 fuel design was selected as the reference fuel design for comparison, consistent with Amendment 22 criteria. GNF2 is also more stable than GE6/7 (P8x8R) fuel as reported in Section 3.9 of Reference 21. 5.0 Mechanical Design and Evaluation 5.1 Fuel Mechanical Design 5.1.1 Design Bases NMP Unit 1 UFSAR Section IV IV-16 Rev. 25, October 2017To meet the performance objectives "2" and "3", the fuel rod is designed with adequate margin to assure that excessive fuel failures will not occur during normal operation or anticipated operational transients. Fuel failure is defined as perforation of the fuel cladding which would permit release of fission products to the reactor coolant. Details of the fuel design can be found in References 7 and 21. 5.1.2 Fuel Rods The reactor fuel consists of high-density ceramic uranium dioxide pellets, manufactured by compacting and sintering uranium dioxide powder into right cylindrical pellets with flat ends and chamfered edges. The pellets are enclosed in Zircaloy-2 tubes which are evacuated, backfilled with helium, and sealed by welding Zircaloy plugs into each end. Ceramic uranium dioxide is chemically inert to the cladding at operating temperatures and is resistant to attack by water. Several U-235 enrichments are used in the fuel assemblies to reduce the local peak-to-average fuel rod power ratios. Selected fuel rods within each reload bundle also incorporate small amounts of gadolinium as burnable poison. Gd2O3 is uniformly distributed in the UO2 pellet and forms a solid solution. The fuel rods are described in further detail in References 3 and 7. The fuel cladding thickness is adequate to satisfy the requirement that the clad be "freestanding" and capable of withstanding pressures well beyond operating reactor pressure without collapsing onto the contained pellets. Adequate free volume is provided within each fuel rod in the form of a pellet-to-cladding gap, and a plenum region at the top of the fuel rod to accommodate thermal and irradiation expansion of the UO2, and the internal pressures resulting from the helium fill gas, impurities and gaseous fission products liberated over the design life of the fuel. A plenum spring or retainer is provided in the plenum space to minimize movement of the fuel column inside the fuel rod during shipping and handling. In barrier fuel, the fuel cladding incorporates an inner lining of pure zirconium: this lining decreases the probability of pellet-clad-interaction induced fuel failures without altering the other properties of the fuel. Two types of fuel rods are used in a fuel bundle: tie-rods and standard rods. Tie-rods are described in Section 5.1.4. The end plugs of the standard rods have shanks which fit into bosses in the tie-plates. An Inconel-X expansion spring is located NMP Unit 1 UFSAR Section IV IV-17 Rev. 25, October 2017over the upper end plug shank of each rod in the assembly to keep the rods seated in the lower tie-plate while allowing independent axial expansion by sliding within the holes of the upper tie-plate. 5.1.3 Water Rods The water rods are hollow Zircaloy tubes with several holes punched around the circumference near each end to allow coolant to flow through. 5.1.4 Fuel Assemblies The fuel assemblies are described in detail in References 3, 8, 19 and 21. Fuel bundle specific information is provided in References 7 and 23. 5.1.5 Mechanical Design Limits and Stress Analysis The fuel mechanical design limits and stress analysis are described in detail in References 3 and 20. 5.1.6 Relationship Between Fuel Design Limits and Fuel Damage Limits Fuel is designed to satisfy the conservative mechanical design limits in accordance with References 3 and 20. 5.1.7 Surveillance and Testing Rigid quality control requirements are enforced at every stage of fuel manufacturing to assure that design specifications are met. Written manufacturing procedures and quality control plans define the steps in the manufacturing process as described in Reference 3. 6.0 Control Rod Mechanical Design and Evaluation 6.1 Design 6.1.1 Control Rods and Drives The control rod drive (CRD) system consists of the control rod, CRD, hydraulic scram system and the hydraulic drive system. The mechanical design and evaluation of the control rod is discussed in this section; the other portions of the control rod system are discussed in Section X-C. NMP Unit 1 UFSAR Section IV IV-18 Rev. 25, October 2017 The cruciform-shaped control rods contain hafnium metal rods or strips and/or a number of vertical stainless steel tubes filled with boron carbide (B4C) powder, compacted to approximately 70 percent of theoretical density. A typical BWR-2 and BWR-6 control rod blade is shown on Figure IV-4 and Figure IV-4a, respectively. Positioning of the control rods relative to fuel assemblies is shown in Reference 3. The relative positioning of the control rod does not change for the different fuel types that comprise the core. An overall plan view of the core showing the position of control rods, fuel assemblies and the in-core nuclear instrumentation is shown on Figure VIII-11. Plugs are welded into the ends of the tubes to seal them. The B4C powder is separated longitudinally into independent compartments by stainless steel balls at approximately 18-in intervals, held in place by a slight swaging of the tube. This feature tends to uniformly spread any compaction of the powder during control rod life. A free volume of approximately 30 percent is provided in each tube as a plenum for helium from the B-10 (n, alpha) Li-7 reaction. The absorber material is held in a cruciform array by a stainless steel sheath extending the full length of the absorber section. A cruciform-shaped top handle aligns the tubes and provides structural rigidity at the top of the control rod. Rollers attached to the top casting maintain the spacing between the control rod and the fuel assembly channels. A similar cruciform connector is located at the bottom of the control rod and contains a velocity limiter section and rollers to position the lower part of the control rod in the control rod guide tube, located below the core. These bottom rollers always remain in the guide tube during operation. A coupling at the bottom of the control rod is connected and locked to the CRD index tube by an expandable ball and socket joint. Design stress intensity limits for control rod poison tubes are given in the following tabulation: Stress Intensity Limits in Terms of ASME Section III-1968 Yield Strength Ultimate Tensile Stress Categories (Sy) Strength (Su) General Primary Membrane 2/3 Sy 1/2 Su Stress Intensity Local Primary Membrane Sy 3/4 Su Stress Intensity NMP Unit 1 UFSAR Section IV IV-19 Rev. 25, October 2017Primary Membrane Plus Sy 3/4 Su Bending Stress Intensity Primary Plus Secondary 2 Sy 1.5 Su Stress Intensity Control rods are internally supported and can withstand external pressures far in excess of that established as the safety limit for the reactor vessel. Advanced control rod designs with an extended lifetime are also used (advanced long life control rod [ALLCR], BWR-6, Duralife 230 [D-230], and Marathon). The advanced control rods have some or all of the following design features: 1. In the upper 6 in of the control rod blade and in the outer edge of each wing, B4C has been replaced by hafnium. Hafnium is a metal and does not expand upon neutron absorption. The regions in which it is used are the highest local neutron absorption regions in the blade. 2. The tubes that house the B4C are made with high-purity type 304 stainless steel, which has improved resistance to cracking. 3. To compensate for the heavier hafnium, the velocity limiter has been redesigned. The total weight of the improved control rod is slightly less than the original control rod, which does not adversely affect scram performance. 4. The BWR/6 control rod coupling release handle design was incorporated. This design is compatible with all existing handling equipment and NSSS hardware. 5. The pin and roller at the tip of each wing are made of PH13-8Mo alloy and Inconel X-750, respectively, to reduce the amount of cobalt in the control rod. The BWR-6, D-230, and Marathon blades have a larger volume of B4C than the ALLCR, which provides an increased useful lifetime. The ALLCR and D-230 control rod generic safety evaluations were approved by the NRC in References 9, 10 and 11. The difference between the Marathon control rod design and the other control rod designs used at NMP1 is the construction of NMP Unit 1 UFSAR Section IV IV-20 Rev. 25, October 2017the absorber zone. The Marathon control rod absorber zone is made of square tubes which are welded together to form the control rod wing. The hollow cylindrical center of the tubes is filled with either hafnium rods or B4C contained in capsules made of 304 stainless steel. The capsule design prevents migration of B4C within the absorber tube and replaces the use of the spacer balls. The design limit for the Marathon control rod absorber tubes is one half the burst pressure for the tube. The Marathon control rod assembly generic safety evaluation report was approved by the NRC in Reference 16. The CRDs are of the locking piston type. A schematic diagram of the drive and hydraulic system is shown on Figure IV-6, and an assembly drawing of the drive mechanism is shown on Figures IV-7 and IV-7A. The drive mechanisms are mounted vertically in thimbles which are welded into the reactor bottom head penetrations. The low end of each thimble terminates in a special flange which contains ports for attaching the hydraulic system lines, and a machined face which mates with a corresponding flange at the lower end of the drive. The operating principles of the control rod hydraulic system are described in Section X-C. At the top end of the drive index tube (the movable element), a coupling is provided which engages and locks into a socket at the base of the control rod as shown on Figure IV-8. The weight of the control rod alone can engage and lock this coupling. Once locked, the drive and rod form an integral unit which must be manually unlocked by specific procedures before a drive or rod can be removed from the reactor. These procedures are established to prevent accidental separation of the control rod from the CRD. The drives position the control rods in 6-in increments of stroke and hold them in these discrete latch positions until actuated for movement by the hydraulic system to a new position. Visible indication of the position of each drive is displayed in the control room by means of illuminated numerals which correspond with the respective latched positions. In addition, indication is provided that shows insert and withdraw travel limits of the drive and an overtravel withdraw limit on the drive have been reached. Control rod seating at the lower end of the stroke prevents the overtravel withdraw limit from being reached unless the control rod is uncoupled from the drive. This allows the coupling to be checked. These indicators and those for the in-core monitors are grouped together and displayed on the control panel and arranged on the board to NMP Unit 1 UFSAR Section IV IV-21 Rev. 25, October 2017correspond to relative rod and in-core monitor positions in the core. During reactor shutdown, the SDM can be verified. The SDM demonstration is performed as described in the Technical Specifications. 6.1.2 Standby Liquid Poison System This system is described in detail in Section VII-C. The standby liquid poison system is designed to provide the capability of bringing the reactor, at any time in a cycle, from a full power and minimum control rod inventory (defined to be at peak xenon) to a cold, xenon-free subcritical condition assuming none of the control rods can be inserted. The liquid poison solution is sodium pentaborate enriched in boron-10 isotope. To meet the shutdown objective, the system is designed to inject a quantity of boron which produces a concentration of at least 109.8 ppm of boron-10 isotope in the reactor core. Cycle-specific liquid poison system SDM results are provided in the SRLR(1). 6.2 Control System Evaluation 6.2.1 Rod Withdrawal Errors Evaluation Design features provided to minimize the possibility of inadvertent continuous control rod withdrawal, and to limit potential power transients in the event they should occur, include the following: 1. The control system is designed so that only one rod can be withdrawn at a time. 2. Normal rod operation is a step (notch) at a time. Two control switches must be operated at the same time to withdraw a rod continuously. 3. The continuous rod withdrawal rate is limited by the control rod hydraulic flow system to a nominal 3 ips. Rod withdrawal rates have been analyzed up to 5 ips, and have been shown not to impact the RWE transient enough to make it a limiting transient. If the withdrawal rate of any control rod is to be left faster than 3 ips, procedural controls shall be implemented to reduce the impact of the fast withdraw time on the RWE transient. NMP Unit 1 UFSAR Section IV IV-22 Rev. 25, October 2017 4. Interlocks prevent rod withdrawal if the neutron flux monitors are not in a condition to provide the required protection, or if the rod withdrawal timing relay should fail. 5. Preplanned withdrawal patterns and procedural controls are used to prevent abnormal configurations giving high rod worths. In the event of a rod sticking in the core during the withdrawal and subsequently dropping out, these procedures will result in excursions no greater than 280 cal/gm. 6. A control RWM will usually be in service backing up these procedural controls. 7. Intermediate and power range level scrams limit power excursions from low reactor power levels. During power operation, in-core monitor alarms warn the Operator if local neutron flux levels approach preset limits. 8. An APRM rod blocking system limits withdrawal of any single control rod which could result in clad damage. NOTE: Operation in IRM range 10 requires total recirculation flow greater than 30 percent of rated during control rod withdrawal. This requirement is based on an analysis of the RWE transient which does not credit the average power range monitor (APRM) rod block system to mitigate this event. If the reload RWE transient analysis requires the APRM rod block system to minimize the CPR for this event, then the validity of the 30-percent flow restriction must be reviewed and dispositioned. A Technical Specification change could be required if the analysis cannot demonstrate that 30-percent flow provides adequate margin. (See GE Report GENE-909-39-1093.) The consequences of inadvertent continuous single rod withdrawal from various power levels is analyzed in Section XV. The consequences of a control rod drop accident (CRDA) resulting in fuel clad failure and release of fission products is also discussed in Section XV. 6.2.2 Overall Control System Evaluation NMP Unit 1 UFSAR Section IV IV-23 Rev. 25, October 2017In order to meet performance objective "6," the core in its maximum reactivity condition must be subcritical with the control rod of highest worth fully withdrawn and all operable rods fully inserted. This criterion has been established to permit withdrawal of a single rod from the core for maintenance purposes. Detailed analysis has demonstrated that several rods in a scatter pattern may be out of the core and a keff less than 1.0 still be easily maintained. In order to assure that the basic criterion will be satisfied, an additional design margin was adopted: that the keff be less than 0.99 with the rod of highest worth fully withdrawn. From an operating point of view, it is only necessary that keff is less than 1.0 with the rod of highest worth fully withdrawn. This limit allows control rod testing at any time in core life and assures that the reactor can be shut down by control rods alone. This limit applies any time in core life. In addition to the standby control rod shutdown requirements, the liquid poison system can shut the reactor down at any time in core life (see Section VII-C). Mechanical malfunctions of CRDs, such that one or more are incapable of insertion, define the requirements for the standby liquid poison system. This system constitutes a redundant, continuously available shutdown capability. The most severe requirement imposed on the standby liquid poison system and its design basis is to shut down from a full-power operating condition, assuming complete failure of the control rod system to respond to an insertion signal. The rate of reactivity compensation provided by the backup system is designed to exceed the rate of reactivity gain associated with reactor cooldown from the full-power condition. This system satisfies performance objective "7." The reactivity control system is designed such that under conditions of normal operation, means are provided for continuous regulation of the core excess reactivity and reactivity distribution. The movement of control rods must not perturb the reactor beyond the capability of an Operator to respond to the disturbance. This requirement prevents unnecessary operation of the reactor protection system (RPS). The maximum rate at which the rods can be moved, and the incremental distance between control drive notches, is such that under normal operating conditions a single notch increment of control withdrawn at the maximum withdrawal rate will result in a stable reactor period of not less than 20 sec. NMP Unit 1 UFSAR Section IV IV-24 Rev. 25, October 2017Under conditions of expected abnormal reactor system disturbances, the reactivity control system provides a sufficient rate of negative reactivity insertion, upon signal of the RPS, to prevent fuel damage. Expected abnormal reactivity disturbances and resulting power transients in the core can derive from any of these three sources. 1. Reactor system induced disturbances of core parameters such as coolant flow or pressure. 2. Single Operator errors or procedural violations. 3. Single equipment malfunctions. The design philosophy for the safety and control systems requires that, for those accidental power transients due to control rod movement with the potential for endangering the health and safety of the public, the system provides, in addition to containment, at least a double level of automatic or inherent protection. For example, CRD housing supports are provided to preclude a control rod shootout, velocity limiters to minimize effects of a control rod dropout and procedural controls to prevent setting up high rod worths. For other incidents, at least a single level of automatic or inherent protection is provided. Analysis of various transients (Section XV) has shown that the negative reactivity insertion rates available lead to rapid termination of all transients. The faster scram times are typically dictated by the rapid isolation transients such as turbine stop valve trip with failure of the bypass valves. For this transient (Section XV), a rapid scram time mitigates the pressure transient, lessening the time that the solenoid-actuated relief valves remain open and maintaining pressure below the setpoint of the safety valves. Therefore, the response of the RPS in combination with the size, heat transfer features and inherent dynamic response characteristics of the core, prevent fuel damage resulting from a reactivity insertion accident due to any single equipment malfunction or Operator error. The inherent safety features of the reactor design, in combination with engineered safeguards such as the control rod velocity limiter and the control rod housing support, are such that the consequences of a potential nuclear excursion accident, caused by any single component failure within the reactivity NMP Unit 1 UFSAR Section IV IV-25 Rev. 25, October 2017control system itself, does not result in damage to the reactor primary coolant system or impair operation of required safeguards. These features and safeguards thus limit potential power excursions such that performance objective "5" is satisfied. Certain postulated rapid reactivity insertion accidents are evaluated for the express purpose of determining whether damage to the primary system would result. To damage the primary system from pressure and momentum effects, rapid (few millisecond) power excursions resulting in rapid fuel rod rupture and dispersion of some quantity of hot fuel into the surrounding coolant would be necessary. This reactor is designed on the basis of avoiding sudden rupture of a significant number of fuel rods in any accidental excursion resulting from component or procedural failure within the reactivity control system. The threshold for this type of fuel rupture is estimated to correspond to a fuel energy content of 425 calories/gm U02. The magnitude of an accidental nuclear excursion is limited first by the strong, negative Doppler coefficient of reactivity inherent in the reactor design, and secondly, by the rate at which positive reactivity is added to the system by the malfunction. Reactivity insertion rates are held below values which could cause primary system damage by the maximum worth that an individual rod can assume and the maximum rate at which it can drop from the core. The rod velocity limiter limits the rate that a rod can fall from the core, and operating procedures normally backed up by the RWM limit the maximum control rod worth. As a design base, reactivity addition rates are limited by these devices to well below the value which could result in any significant amount of fuel reaching the sudden clad rupture range during an accidental excursion. The nominal value for excursion consequences used in setting control system and engineered safeguards design is 280 cal/gm. 6.3 Limiting Conditions for Operation and Surveillance As a result of the foregoing physics analysis and mechanical evaluations in this section, a series of limiting conditions for operation (LCO) and surveillance requirements are provided for the CRD system. These include items such as shutdown margins, coupling integrity, individual rod worth and scram times. Suitable surveillance requirements are established to cover NMP Unit 1 UFSAR Section IV IV-26 Rev. 25, October 2017testing and test frequency. The LCO and the surveillance requirements are described in the Technical Specifications. 6.4 Control Rod Lifetime Control rod lifetime is governed by both nuclear and mechanical qualities. The principal absorber material in the control rod is B10. The nuclear reaction governing the absorption phenomena of B10 is the () reaction: B10 + n --> He4 + Li7 The limiting mechanism on control rod lifetime has been determined to be the stress corrosion cracking of the stainless steel tubing induced by sintering of B4C particles and the resultant swelling of the compacted B4C as helium and lithium concentrations grow. The cracks propagate through the tube wall allowing reactor coolant to enter the tube. The subsequent leaching of the B4C out of the tubing results in a loss of B4C and, therefore, a reduction of control blade effectiveness faster than that which would be predicted for neutron absorption alone. The end of control rod design life has been defined as that point at which a 10-percent reduction in relative control rod worth is reached, which corresponds to 34-percent B10 depletion averaged over any quarter segment of all B4C control rod designs. The safety significance of boron loss is its impact on shutdown capability and scram reactivity. In order to assure that the aforementioned impact is negligible, a control blade changeout program is provided which assures that all B4C control rod designs do not exceed the lifetime limit of 34-percent B10 depletion averaged over any 1/4 of the blade. This program is in compliance with the preferred action contained in Inspection and Enforcement Bulletin 79-26. In addition to the control rod changeout program, a shutdown margin test is performed (as part of the startup tests) in accordance with Technical Specifications to demonstrate that the reactor can be made subcritical (this is based on the premise that the calculated control rod worths used in the test are based on the assumption that no boron loss has occurred). The extended lifetime of the ALLCR, D-230, and Marathon control rod designs is achieved by eliminating cracking of the B4C tubes. Hafnium is used where neutron exposure is highest (on the tip of the blade and on the outer edge of each wing), to reduce peak NMP Unit 1 UFSAR Section IV IV-27 Rev. 25, October 2017local neutron depletion in the B4C tubes. The absorber worth of hafnium also decreases slower than B4C. 7.0 Reactor Vessel Internal Structure 7.1 Design Bases The reactor internals are designed to withstand the design basis earthquake (DBE), thereby meeting performance objective "1." The design of internal components of the reactor vessel (in conjunction with the reactor vessel) allows adequate core cooling to be maintained during both normal operation and accident conditions without failure (see Section XVI-A.2.7). The internal components are designed to: 1. Provide support for the fuel, steam separators, dryers, etc., during normal operation and accident condition. 2. Maintain required fuel configurations and clearances during normal operation and accident conditions. 3. Circulate reactor coolant to cool the fuel. 4. Provide adequate separation of steam from water. Stresses in various core components are at maximum during the blowdown resulting from the main steam line (MSL) break discussed in Section XV. 7.1.1 Core Shroud The core shroud, as shown on Figure IV-9, is a stainless steel cylinder which surrounds the core and provides a barrier to separate the upward flow of coolant through the core from the downcomer recirculation flow. Mounted at the top of the shroud is the shroud headsteam separator assembly. A discharge plenum at the top of the core provides a mixing chamber before the steam-water mixture enters the steam separators. The recirculation inlet and outlet plenums are separated by shroud and shroud support. The shroud support is designed to sustain the differential expansion of the ferritic reactor vessel and the austenitic stainless steel shroud without high stresses. The shroud support is fabricated from solid Inconel. The shroud support essentially sustains all of the vertical weight of the core structure (except the fuel assembly weights transmitted to NMP Unit 1 UFSAR Section IV IV-28 Rev. 25, October 2017the guide tube) and the steam separator assembly; the differential upward pressure loading on the shroud under operating conditions; and the vertical and sidewise thrusts developed on the core and core structure during an earthquake. The cylindrical shroud is joined to the shroud support with a full penetration weld. The shroud support plate, tie-rods, head bolts, and associated welds are fabricated using Inconel stainless steel. The principal stresses produced in the shroud are due to differential pressure loading, differential thermal expansion, deadweight loadings and earthquake loadings. Core Shroud Intergranular Stress Corrosion Cracking (IGSCC) The core shroud vertical and horizontal welds are susceptible to IGSCC as discussed in References 12 and 13, and NRC Generic Letter (GL) 94-03. The core shroud horizontal and vertical welds have been inspected and determined to have IGSCC in and near the heat-affected zone (HAZ) of the welds. This cracking has been evaluated and determined to be prototypical of IGSCC reviewed by the NRC as part of GL 94-03 and addressed by the Boiling Water Reactor Vessel and Internals Project (BWRVIP) core shroud IGSCC documents. The NRC safety evaluation report (SER) for the generic application of the BWRVIP core shroud inspection and evaluation document is applicable to the Nine Mile Point Nuclear Station - Unit 1 (Unit 1) IGSCC. The shroud inservice inspections (ISI) have determined that the horizontal and vertical welds inspected satisfy the required structural margins considering the existing IGSCC, and maintain the core shroud such that all design basis requirements are satisfied. The horizontal welds have core shroud stabilizer assemblies (tie-rods) installed which structurally replace horizontal welds H1 through H7 such that ISI of the horizontal welds is not required. The vertical weld integrity is required considering the core shroud stabilizer design basis assumption of 360-deg throughwall cracking of the horizontal welds. Complete vertical weld throughwall cracking can be tolerated for the vertical welds provided horizontal weld integrity is established by inspection. Since horizontal weld inspections are not performed, vertical weld ISI is required to maintain the core shroud stabilizer design basis assumptions. The required ISI interval for the vertical welds is defined based on the References 12, 13 and 14 approved methods. The specific interval is defined by engineering analysis of the as-found cracking and consideration for potential crack growth and inspection uncertainty. NMP Unit 1 UFSAR Section IV IV-29 Rev. 25, October 2017The primary stress which could cause vertical weld failure results from the internal pressure. Consistent with ASME Code Section XI practice, internal pressure is the only load to be considered for axial cracks. Other loads such as deadweight, seismic and thermal expansion have negligible impact and need not be considered. The design basis internal pressures which define the limiting faulted condition are 22 psi for the upper shroud (H1 through H6A) and 63 psi for the lower shroud (below core plate) (see Table XVI-9). The allowable flaw sizes consider the internal pressures under all conditions: normal, upset and accident. The required ASME Code Section XI safety factors of 3.0 for normal and upset conditions and 1.5 for emergency and faulted conditions are applicable consistent with the ASME Code requirements for evaluating axial flaws. The potential impact of approximately 180 in of throughwall vertical weld crack leakage has been determined to be less than .11 percent of total core flow. The results show that at rated power and core flow the predicted leakage is sufficiently small so that the steam separation system performance, cavitation protection, core monitoring, fuel thermal margin and fuel cycle length remain adequate. Since the core flow leakage is minor and only a postulated condition, no core monitoring correction should be applied or is required. Also, this leakage flow has no impact on Section XV LOCA analyses since the core cooling function is performed by core spray cooling, not reflood, and, therefore, leakage from the shroud to the annulus region has no effect on core cooling. The core shroud was reinspected during refueling outage (RFO) 15. A preemptive repair of the V9 and V10 welds was performed during RFO15 by installing a contingency repair clamp design previously approved by the NRC. The vertical weld repair clamps are described in Section IV-B.7.1.10. The vertical repair clamps replace the load carrying function of the V9 and V10 welds; therefore, future inspections of the V9 and V10 welds are not required. 7.1.2 Core Support A 304 stainless steel fuel support casting is mounted on top of each control rod guide tube. Each guide tube, with its fuel support casting, bears the weight of four fuel assemblies and rests on a CRD housing welded to the stub tube mounted on the vessel bottom head. The core plate provides lateral guidance for the bottom of the fuel assemblies. Each casting contains four orificed flow passages, one for each of the four fuel NMP Unit 1 UFSAR Section IV IV-30 Rev. 25, October 2017assemblies. The orifice regulates the core flow through the fuel assemblies. The CRD housing is welded to a stub tube and the stub tube is welded to the vessel bottom head. The stub tube provides a portion of the reactor vessel pressure boundary and provides support to the CRD housing. IGSCC in the stub tubes has been identified due to the 304 stainless steel stub tube being furnace sensitized during the original vessel fabrication post-weld heat treatment. Cracking is throughwall as made evident by leakage detected from the CRD penetrations during under vessel leakage inspections. The leakage is eliminated by an ASME-approved roll expansion process. The roll expansion process creates an interference fit at the interfacing surfaces of the CRD housing, vessel bottom head bore and stub tube. The roll expansion process has been demonstrated and qualified by analysis and testing to eliminate leakage and provide structural support with postulated 360-degree throughwall cracking in the stub tube. The use of the ASME-approved roll expansion process has been approved for use at Nine Mile Point by the NRC in Reference 22. 7.1.3 Top Grid The upper core grid provides lateral support and alignment at the top of the fuel assemblies with four fuel assemblies contained in each grid opening. The grid assembly is supported from the core shroud. 7.1.4 Control Rod Guide Tubes The stainless steel control rod guide tubes extend from the CRD housings through holes in the core plate. Each tube is designed as a lateral guide for a control rod and as vertical support for the four fuel assemblies surrounding the control rod. In addition, the guide tubes protect withdrawn control rods from cross flow in the inlet plenum region. The downward vertical loads from the fuel assemblies are directly transferred to the guide tubes and to the bottom vessel head. The guide tubes are locked into a sleeve mounted on the CRD housing. This sleeve is inserted in the CRD housing and extends from the housing flange almost the entire housing length. This locking device prevents upward movement of the guide tube. The guide tubes are constructed of 10-in 304 stainless steel pipe. 7.1.5 Feedwater Sparger NMP Unit 1 UFSAR Section IV IV-31 Rev. 25, October 2017The feedwater sparger is mounted to the reactor vessel wall above the downcomer annulus formed by the shroud and vessel. The sparger discharges water radially inward. This arrangement permits the cooler feedwater to mix with the downcomer recirculation flow before coming in contact with the reactor vessel. This mixing also minimizes carry-under and increases the recirculation pump suction subcooling. 7.1.6 Core Spray Spargers The core spray spargers with spray nozzles are mounted along the inside of the core shroud in the discharge plenum at the top of the core. A more detailed description is given in Section VII-A. 7.1.7 Liquid Poison Sparger The ring sparger for the injection of liquid neutron absorber is mounted on the inside shroud surface below the core. 7.1.8 Steam Separator and Dryer The steam separator assembly consists of a base into which are welded an array of standpipes, with a steam separator located at the top of each standpipe. The steam separator base assembly forms the shroud head which is the top of the core discharge plenum. The fixed centrifugal-type steam separators have no moving parts. In each separator, the steam-water mixture rising through the standpipe impinges on vanes which impart a spin to establish a vortex which separates the steam from the water. The separated water enters the pool that surrounds the standpipes to enter the downcomer annulus. The steam dryer assembly is mounted to the reactor vessel wall above the separator assembly. A shroud extends from the bottom of the dryer assembly below the pool surrounding the standpipes to form a seal separating the fluid entering and leaving the dryer. Steam from the separators flows upward and outward through the drying structures. Moisture removed by the dryers flows through a system of troughs and drain tubes, the drain tubes extending below the pool surrounding the standpipes. Dry steam enters the vessel head cavity and is directed into the steam outlet nozzles. Vertical guide rods on the inside of the vessel provide a guide train for the dryer assembly and shroud head when being installed. The steam separator and the dryer NMP Unit 1 UFSAR Section IV IV-32 Rev. 25, October 2017assembly are bolted to the core shroud flange by long holddown bolts that extend above the separator and dryer for easy access. The separator base is guided to its final position with locating pins. Vertical track guides are used to rough position the shroud head. 7.1.9 Core Shroud Stabilizers The core shroud structure was fabricated by welding. Core shroud weld numbers H1 through H6B are all of the horizontal (circumferential) shroud welds. Core shroud weld numbers V1 through V16 are all of the vertical and top guide ring segment welds. Weld H7 attaches the shroud to the forged stainless steel shroud support ring. Weld H8 is a bimetallic weld that attaches the stainless steel support ring to the Inconel core support cone. Ultrasonic and enhanced visual examinations of weld H8 confirmed the weld is structurally adequate. The core shroud stabilizers are designed to structurally replace shroud welds H1 through H7 in the event that the shroud welds are cracked. The tie-rod assemblies combined with core plate wedges replace welds H1 through H7. The shroud repair hardware, by design, was not intended to structurally replace the shroud vertical/segment welds. A more detailed description of the shroud stabilizer design is provided in Section XVI-A. Tie-Rod Assembly Description Each tie-rod assembly consists of a tie-rod, upper support, upper spring, middle support, lower lateral and axial springs, lower support with two toggle bolts, and other minor components. The ends of the tie-rod assemblies are attached at the top to the upper shroud head flange and at the bottom to the Inconel shroud conical support. The shroud head is notched at four azimuth locations (eight notches) using electric discharge machining (EDM) to accommodate the installation of the upper stabilizer support. At the bottom, two holes are machined through the angled conical support for attaching each tie-rod assembly. The tie-rod assemblies are designed to prevent unacceptable lateral or vertical motion of the shroud shell sections, assuming complete failure (360-deg throughwall) of one or more of the circumferential shroud welds. Each cylindrical shell and ring section of the shroud is prevented from unacceptable motion by the stabilizers. The functions of each tie-rod assembly component are as follows: NMP Unit 1 UFSAR Section IV IV-33 Rev. 25, October 2017 1. The tie-rods serve to provide an alternative vertical load path from the upper support of the tie-rod assembly through the shroud support cone. These tie-rod assemblies maintain the alignment of the core shroud to the reactor vessel. 2. The upper support bracket combined with the upper lateral spring are designed to restrain lateral movement of the shell between welds H1 and H2, the ring between H2 and H3 and the shell between H3 and H4. 3. The lateral rigid support (limit stop) located at the midpoint of the tie-rods is designed to restrain lateral movement of the shell between welds H4 and H5. The rigid support is also provided for the tie-rod so that the tie-rod's natural frequency will be higher than that of the forcing frequency due to flow-induced vibration. 4. The lower lateral spring contacts the shroud and the reactor pressure vessel (RPV) and is designed to restrain lateral movement of the shell between welds H5 and H6A via the core plate bolts and wedges, the ring between welds H6A and H6B, and the shell between H6B and H7. It is noted that the original installation of the lower lateral spring contacts on each tie-rod assembly did not capture shroud weld H6A as was intended by design. The lower spring contacts were subsequently modified to extend beyond weld H6A. 5. The lower axial spring is designed to provide axial flexibility of the tie-rods to accommodate postulated temperature transients. 6. The lower support with toggle bolts is designed to provide an attachment of the tie-rod assemblies to the shroud conical support and to minimize leakage between the RPV lower plenum inlet flow and the RPV annulus flow. Core Plate Wedge Description The shroud stabilizers also include four core plate wedges (spacers) located in the annulus between the core support plate and the inside of the shroud. In the event that welds H6A and NMP Unit 1 UFSAR Section IV IV-34 Rev. 25, October 2017H6B failed, the wedges would provide a direct load path from the core plate to the shroud to help distribute the lateral loads occurring during a seismic event. The shroud ring at this location is restrained in the lateral direction by the lower tie-rod lateral spring. The wedges are held in place by clamping against the existing angle brackets that position the existing shield blocks. 7.1.10 Core Shroud Vertical Weld Repair The core shroud vertical weld repair addresses the cracking of vertical welds V9 and V10. The repair basically consists of a clamp with a plate with attached pins which are inserted into holes which are machined by the EDM process on either side of the flawed vertical weld. The clamp bridges across the flawed vertical weld and transmits the pressure load normally transmitted through the vertical weld. Two clamps are used for the V9 weld and two clamps for the V10 weld. The repair was designed and installed under Reference 15. Refer to Section XVI-A.5.2 for additional vertical weld repair design information. 7.2 Design Evaluation The core structural components are designed to accommodate the loadings applied during normal operation and maneuvering transients, considering both stress and deflection. Deflections are limited so that the normal functioning of the components under these conditions will not be impaired. Where deflection is not the limiting factor, the ASME Boiler and Pressure Vessel Code, Section III-1968, is used as a guide to determine limiting stress intensities and cyclic loadings for the core internal structure. The reactor internals are designed to preclude failure which would result in any part being discharged through the steam line in the event of a steam line break beyond the flow restrictors outside of the steam line isolation valve. The structural components which guide the control rods have been examined to determine the loadings which would occur in a LOCA (including a steam line break). The core structural components are designed so that deformations produced by accident loadings will not prevent insertion of control rods. Considerable effort was expended to eliminate possible failures or control instability due to the vibration of reactor internal NMP Unit 1 UFSAR Section IV IV-35 Rev. 25, October 2017components. The reactor system was analyzed as a multidegree-of-freedom system. This analysis determined the system's natural frequencies, the resultant vibration mode shapes and the relationship between the vibration amplitudes and the critical stresses in the system, to show that system integrity would be maintained. 7.3 Surveillance and Testing Rigid quality control requirements assured that the design specifications of the vessel internal components were met. These quality control methods were utilized during the fabrication of the individual components as well as during the assembly process. Preoperational performance tests and the startup program demonstrated the design adequacy of reactor vessel internals and operability of the core spray spargers. Periodic testing of the control rod system, i.e., reactivity margin - core loading and stuck control rods; rod scram insertion times and reactivity anomalies, is described in the Technical Specification. C. REFERENCES 1. "Supplemental Reload Licensing Report for Nine Mile Point 1 Reload 24 Cycle 25," 002N6949, Revision 0, March 2017. 2. Randall and St. John, Nucleonics 16(3), 82-86, 129 (1958). 3. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-24-US, March 2017. 4. General Electric Company, "Steady-State Nuclear Methods," NEDE-30130-P-A, April 1985. 5. "Qualification of the One-Dimensional Core Transient Model for BWR's," NEDO-24154, Vol. 1 and 2, and NEDE-24154-P-A, Vol. 3, February 1, 1986. 6. R. B. Linford, "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NEDO-10802, December 1986, and Amendments. 7. "Global Nuclear Fuels Fuel Bundle Designs," NEDE-31152P, Revision 9, May 2007. NMP Unit 1 UFSAR Section IV IV-36 Rev. 25, October 2017 8. GENE-770-31-1292, Rev. 2, "Engineering Report for Application of GE11 to NMP1 Reload 12," General Electric Co. Proprietary Document, April 1993. 9. Safety Evaluation of the General Electric Advanced Long Life Control Rod Assembly, NEDE-22290, January 1985. 10. Letter from C. O. Thomas (NRC) to J. F. Klapproth (GE), July 1, 1985. 11. Safety Evaluation of the General Electric Duralife 230 Control Rod Assembly, NEDE-22290-P-A Supplement 3, May 1988. 12. GENE-523-113-0894, Rev. 1, "BWR Core Shroud Inspection and Evaluation Guidelines," March 1995. 13. BWRVIP-01, Rev. 2, "BWR Core Shroud Inspection and Flaw Evaluation Guideline," October 1996. 14. BWRVIP-07, "Guidelines for Reinspection of BWR Core Shrouds," February 1996. 15. Design Change No. N1-97-033, "Core Shroud Vertical Weld Contingency Repair." 16. GE Marathon Control Rod Assembly, NEDE-31758P-A, October 1991. 17. NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," August 1999. 18. NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," August 1999. 19. GEH-0000-0118-3592-R1, "GNF2 Fuel Design Cycle-Independent Analyses for Constellation Energy Nuclear Group Nine Mile Point Nuclear Station Unit 1," March 2011. 20. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-24, March 2017. 21. "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II)," NEDC-33270P, Revision 7, October 2016. 22. Letter from N.L. Salgado (NRC) to S.L. Belcher (NMPNS), dated August 3, 2009, Nine Mile Point Nuclear Station, Unit NMP Unit 1 UFSAR Section IV IV-37 Rev. 25, October 2017No. 1 - Request to Utilize the Alternative of Applying ASME Code Case N-730 for the Repair and Inservice Inspection of Control Rod Drive Bottom Head Penetrations for the License Renewal Period of Extended Operation (TAC No. MD9604). 23. "Fuel Bundle Information Report for Nine Mile Point 1 Reload 24 Cycle 25," 002N6950, Revision 0, March 2017. LIMITING POWER/FLOW LINE (Typical) &A.I 2 ::; 0 0 _, L&. CIO ....... °' w 0 Q. CJ) 2: 0 .... (,0 0 *V .... ----.----------..-------------------..... 0 0 0 co 0 <O 0 v 0 N :JO %) 1VVHt3H.1 RGUMIV-1 UFMlt -..W. M *&.Mte 1-t :: 0 .... L&. w Qc 0 u 0 MJ .... <( ex .... :z '41 u ex ""' Q. Nine Mile Point Unit 1 FSAR FIGURE IV-2 THRU FIGURE IV-3 FIGURES IV-2 THRU IV-3 HAVE BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 TYPICAL BWR-2 CONTROL ROD -ISOMETNC 6.5" SHEATH * * ? ; : . ; : j .* . 9.75" COUPLING RELEASE HANDLE FIGURE IV-4 UFSAR Rev. 15 (Nov. 1997) TYPICAL MODilflED BWRI CONTROL BLADE -1.SQ.M,ETR.IC I .. . ' . ** I \'

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Nine Mile Point Unit 1 FSAR FIGURE IV-5 THIS FIGURE HAS BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 CONTROL ROD DRIVE AND HYDRAULIC SYSTEM CONTROL BLADE ACCUMULATOR _/ INLET SCRAM VALVE BALL OiECK VflL.VE OiECK VALVE COOLING WATER HEADER WITHDRAW VALVES INDEX TUBE MAIN DRIVE PISTON 8 SEALS EXHAUST SCRAM VALVE EXHAUST HEADER FIGURI IV-8 UFIAR Rev. 14 (June 1998) Part No. 3: Uncoupling Rod I. llTEMAl.AllO...._Y l. *T£nAL flllElt ASShllt.Y -----l. UllCDUf\.IJlllftOOASSERY *. """"CAI' l.IMltL .. SlOPl'lttl* 1. COi.UT SPRMIG I. COLLET A(llD COLLET PISTON !. COllfTHl'.lllllGthlltlicrlilldlr,W*flafttt) 10. COlltT Ptmll SUU 11. SPAttRl,._.lfcytllllls,*, ... "-') 11. 11Jrn:11 OltlflC£S!TJ'$1itlll U. POSITIOll lltrlltATOlt S'IJTCllEStT,,iCll) It. lOCltlC GROOVE tTJtliUll K. OUTU tBIE (P'lrlll cyM*, Wit, *"-Ill 11.CYl.11111£11TIJI( 11. tl0£X TUIE 11. UIDUIG IAlllD It. *THIW.. MDIII SEAL llJIG$ lf1Jitlll lt. IO!tJllM. "'11* lllllltlOllS !T-1 21. EIT£1111Al.l'ISTOllllUSHlllGS n. llTERllAL PISTI* !£ALI ll. STUllKI 24. COOLllC *TO Ot.-ICE n. DRlvt*llSln IAtU IM.ET {lkmll ... ICl.i*i'IH'illdl'nll*tlll) 25. IALL*SHUTTU YALV£ n. R£ACT01*u**Ln1nr**M._.., BWR-2 CONTROL ROD DRIVE ASSEMBLY 21. SIJTQfoACJIAt.c llKftET '"""tilt pist\11) JO. DRtYE-llTllDllAI POltT$,MID NUIULUS (At..-.uetJ Jl.RllCfl.MCI: ... -sal("llf"'"" U. P0$1TrOll HIDICATDR PROBE 3'. POSITllll ldlCATOfl CABLE Jl.l'ISTOll1'1*""1 3L CAPSCllP!T ..... I 11.IHffllllOU ** DfllVE-llfst'ltT PUlllTS ANO AllJlllLUS 9. Dllvt f\NIGE (,_t atcr1*, hllll. *111111111111 41. UllOCUIC l'OllT""' .Wiii.iii __ .. __ , II. DftM-ltnmAl'IAJU Ill.ET fAISI Mitt Ill su* wal*l 42. fETAL l>R*G SEAL fl'lrivt lo llOuslftll U. OllVEPISTOR '4. INOICAT<ll Tllll (Pal ti'""" hlllel fS, ntllMlCOUPlE (Piii ii 111111"'* HUD 11*t .fi. 47. SPRIMG IASHEllS 41. STOfl PISTOlll IUSHltlGSIT,..ufl '9. -. coun F111GE1t 1T7'it1h SI. COTTER Pll 57. COUPLING SPUD FIGURE IV-7 UFSAR Rev1s1on 16 November Jqqq 2

• SEAL llllG l . FLAT HlAD SCIU (O-Rlllli SPAUR MOUNlllG)(SEE FIG. 2-2) I . S
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• DRIV[*UP S£ALS) 13 -WATll PORTS IR CDLL£T llOUSllOG 74
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_,.,/ TYPICAL CONTROL ROD TO DRIVE COUPLING -ISOMETRIC CONTROL ROD ASSEMBLY wNELOCITY LIMITER ACTUATING SHAFT FIGURE IV-8 UFSAR Rev. 14 (June 1998)

• " ., I I SOURCE* REACTOR VESSEL ISOMETRIC STEAM OUTLET NOZZLE -------i-r-, WATER LEVEL MEASUREMENT TAPS-======= SEPARTION ASSEMBLY -------tff-:t-tt---.i HOLD DOWN BOLT FEEDWATER SPARGER --------;"'11 CORE SPRAY NOZZLE------CORE SPRAY SPARGER UPPER CORE GRID IN-CORE FLUX MONITOR ------#-llHl---t ASSEMBLY CORE SHROUD ----------1+-111 STABILIZER CORE SHROUD ----------tt-91 VESSEL WALL ----------CONTROL ROD GUIDE TUBE RECIRCULATING WATER INLET NOZZLE IN-CORE MONITOR-------------i SLEEVE IV. 9. dgn ( cedd> VESSEL HEAD STUD --------VESSEL FLANGE ------STEAM PRESSURE MEASUREMENT FEEDWATER INLET NOZZLE CORE DIFFERENTIAL PRESSURE TAP ;-----------CONTROL ROD DRIVE HOUSING FIGURE IV-9 UFSAR Rev. 14 June 1996 U.S. NUCLEAR REGULATORY COMMISSION DOCKET 50-220 LICENSE DPR-63 NINE MILE POINT NUCLEAR STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT (UPDATED) OCTOBER 2017 REVISION 25 NMP UNIT 1 UFSAR Section V EF V-1 Rev. 25, October 2017 SECTION V LIST OF EFFECTIVE FIGURES Figure Number Revision Number V-1 19 V-2 16 V-3 14 V-4 16 V-5 16 V-6 16 V-7 16 V-8 16 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section V V-i Rev. 25, October 2017 SECTION V REACTOR COOLANT SYSTEM A. DESIGN BASES 1.0 General 2.0 Performance Objectives 3.0 Design Pressure 4.0 Cyclic Loads (Mechanical and Thermal) 5.0 Codes B. SYSTEM DESIGN AND OPERATION 1.0 General 1.1 Drawings 1.2 Materials of Construction 1.3 Thermal Stresses 1.4 Primary Coolant Leakage 1.5 Coolant Chemistry 2.0 Reactor Vessel 3.0 Reactor Recirculation Loops 4.0 Reactor Steam and Auxiliary Systems Piping 5.0 Relief Devices C. SYSTEM DESIGN EVALUATION 1.0 General 2.0 Pressure 3.0 Design Heatup and Cooldown Rates 4.0 Materials Radiation Exposure 4.1 Pressure-Temperature Limit Curves 4.2 Temperature Limits for Boltup 4.3 Temperature Limits for In-Service System Pressure Tests 4.4 Operating Limits During Heatup, Cooldown, and Core Operation 4.5 Predicted Shift in RTNDT 4.6 Neutron Fluence Calculations 5.0 Mechanical Considerations 5.1 Jet Reaction Forces 5.2 Seismic Forces 6.0 Safety Limits, Limiting Safety Settings and Minimum Conditions for Operation D. TESTS AND INSPECTIONS NMP Unit 1 UFSAR TABLE OF Section Title Section V V-ii Rev. 25, October 2017 1.0 Prestartup Testing 2.0 Inspection and Testing Following Startup 2.1 Pressure Test 2.2 Pressure Vessel Irradiation E. EMERGENCY COOLING SYSTEM 1.0 Design Bases 2.0 System Design and Operation 3.0 Design Evaluation 3.1 Redundancy 3.2 Makeup Water 3.3 System Leaks 3.4 Containment Isolation 4.0 Tests and Inspections 4.1 Prestartup Test 4.2 Subsequent Inspections and Tests F. REFERENCES NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section V V-iii Rev. 25, October 2017 V-1 REACTOR COOLANT SYSTEM DATA V-2 OPERATING CYCLES AND TRANSIENT ANALYSIS RESULTS V-3 FATIGUE RESISTANCE ANALYSIS V-4 CODES FOR SYSTEMS FROM REACTOR VESSEL CONNECTION TO SECOND ISOLATION VALVE V-5 TIME TO AUTOMATIC BLOWDOWN V-6 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section V V-iv Rev. 25, October 2017 V-1 REACTOR EMERGENCY COOLANT SYSTEM V-2 REACTOR VESSEL NOZZLE LOCATION V-3 REACTOR VESSEL SUPPORT V-4 thru FIGURES DELETED V-7 V-8 EMERGENCY CONDENSER SUPPLY ISOLATION VALVES (TYPICAL OF 2)

NMP Unit 1 UFSAR Section V V-1 Rev. 25, October 2017 SECTION V REACTOR COOLANT SYSTEM A. DESIGN BASES 1.0 General The reactor coolant system (RCS) includes the reactor primary system, solenoid-actuated relief valves, primary system safety valves, and the emergency cooling system. The reactor primary system in turn includes the reactor vessel and internals, recirculation piping, valves, pumps and all connected piping to the external containment isolation valve. Section XVI, Special Topical Reports, details the design of the reactor vessel and summarizes many design analysis calculations covering steady-state analyses, pipe reaction calculations, cyclic loading, and summaries of tests performed during fabrication. Surveillance procedures during Station operation are also included. 2.0 Performance Objectives The performance objectives for the RCS are as follows: 1. Seismic considerations are as described in Section III. 2. The RCS is designed to safely accommodate tripping of the turbine generator, loss of power to the reactor recirculation system, and other transients which can be expected from Station operation and normal maneuvers. 3. Serious primary reactor system overpressure is prevented. 4. Power excursions from any credible reactivity addition accident will not rupture the pressure vessel or impair operation of required safeguards equipment. 5. Components of the reactor primary coolant system are operated so that no substantial pressure is imposed unless the reactor vessel materials are 60°F above the nil ductility reference temperature (RTNDT). NMP Unit 1 UFSAR Section V V-2 Rev. 25, October 2017 6. Heat removal systems are provided which are capable of safely accommodating core decay heat following isolation from the main condenser due to auxiliary power loss or loss of coolant from the reactor. Pressures, temperatures and other pertinent design data are presented for all major components in Table V-1. To meet performance objective 2 for moderately frequent events (i.e., main steam isolation valve (MSIV) closure with scram), the reactor is designed to accommodate rapid isolation from the main condenser without the system pressure rising high enough to actuate the safety valves. For infrequent events (i.e., turbine trip without bypass), predicted peak pressures can exceed safety valve setpoint since probability of event initiation is low. The recirculation piping and pumps are designed to accommodate a loss of power such that the resultant flow coastdown will maintain the reactor core minimum critical power ratio (MCPR) the safety limit critical power ratio (SLCPR) (Section IV). To meet performance objective 3, overpressure scram, an emergency cooling system, solenoid-actuated relief valves, safety valves and a turbine bypass system are included. To meet performance objective 4, the maximum reactivity addition rate for the most severe credible accident should not result in a peak fuel enthalpy greater than about 425 cal/gm. As discussed in Section XV, the fuel enthalpy following a postulated control rod dropout is well below this limit. Performance objective 5 is met with margin by careful attention to design, taking into account the change in material properties with irradiation (as discussed in Section V-C.4.0). To meet performance objective 6, an emergency cooling system (Section V-E) is included to remove energy from the reactor system following isolation from the main condenser (normal heat sink), and a core spray system (described in Section VII-A) is included to remove energy from the core following a loss-of-coolant accident (LOCA). 3.0 Design Pressure The nominal operating pressure of the system is 1030 psig. The reactor vessel design pressure of 1250 psig is determined by analysis of margins required to provide a reasonable range for NMP Unit 1 UFSAR Section V V-3 Rev. 25, October 2017 maneuvering during operation, with additional allowances to accommodate transients above the operating pressure without actuation of the safety valves. Analyses presented in Section XV demonstrate the ability of the Station to safely withstand all anticipated disturbances with resultant pressures well below 1250 psig. The design pressures for the piping and other primary system components are based on the reactor vessel design pressure with considerations for static and dynamic heads due to elevation and pump discharge pressure, and overpressure allowances defined in the ASME Boiler and Pressure Vessel Code, Section I-1962 and the ASA B31.1-1955 Piping Code. 4.0 Cyclic Loads (Mechanical and Thermal) Fatigue resistance of the reactor vessel was originally analyzed based on the expected number of operating cycles over the 40-yr life of the vessel. Table V-2 lists the operating cycles evaluated and their expected number of cycles. Using the operating cycles in Table V-2, stress analyses were performed on the feedwater nozzles, control rod drive (CRD) penetrations, lower vessel head, vessel support skirt, core support cone, vessel wall, other nozzles in the vessel, closure studs and the basin seal skirt weld. Fatigue usage factors, utilizing the expected number of operating cycles in Table V-2, were calculated as follows: Where: u = fatigue usage factor ni = expected number of cycles of a given stress amplitude Ni = maximum allowable number of cycles at the same stress amplitude The calculated usage factors (Table V-3) were all within the allowable design limits (General Electric Company (GE) - 0.8, ASME Section III-1965 - 1.0). NMP Unit 1 UFSAR Section V V-4 Rev. 25, October 2017 Except for the reactor recirculation nozzles, stress analysis on other nozzles in the reactor vessel concluded that they were subjected to significantly less severe transients than the feedwater nozzles and, therefore, their fatigue usage factors were all negligible. Stress analyses of recirculation nozzle thermal transients conclude that the nozzle fatigue usages are negligible. The vessel wall was also concluded to have a negligible fatigue usage factor. The above analyses were based on the expected number of operating cycles using assumed parameters for each transient. In addition, NMP has implemented a Fatigue Monitoring Program (FMP) that is used to manage the fatigue life of analyzed components. This program utilizes the FatiguePro software to maintain all actual calculated cumulative usage factors (CUF) below their corresponding allowables. The calculations are based on the actual, rather than expected, number of cycles experienced by the plant for each transient and, in some cases, the actual rather than assumed parameters experienced by each cycle. 5.0 Codes Applicable codes for the RCS are included in Table V-4. Discussion of calculations demonstrating Code adherence are given in Section XV. Further summaries are provided in Section XVI. Codes applicable up to the outside of the second isolation valve on all auxiliary and emergency systems are also given in Table V-4. B. SYSTEM DESIGN AND OPERATION 1.0 General 1.1 Drawings A flow diagram of the RCS is shown on Figure V-1. This system is defined as encompassing the reactor primary system, the solenoid-actuated relief valves, primary system safety valves and the emergency cooling system. The reactor primary system, in turn, includes the reactor vessel and its internals, recirculation piping, valves, pumps and all connected piping to the external containment isolation valve. NMP Unit 1 UFSAR Section V V-5 Rev. 25, October 2017 Many other systems connect directly to the RCS besides those shown on Figure V-1. These are included in separate flow diagrams as given below: System Figure No. Feedwater XI-7 Shutdown Cooling X-1 Cleanup X-2 Core Spray VII-1 Liquid Poison VII-6 1.2 Materials of Construction Insulation throughout the RCS within the drywell consists of metal reflective insulation and blanket insulation. In the event of small coolant leakage leading to wetting of the insulation in contact with the outer surface material of the loop, no adverse electrochemical or chemical reaction leading to excessive corrosion is anticipated. 1.3 Thermal Stresses Heatup and cooldown rates for water in the reactor system during normal operation will be limited to 100°F/hr by procedural control. Holding this limit will assure that stresses are well within Code limits as discussed in Section V-A.4.0 above. In the event of a short-term, more rapid blowdown greater than 100°F/hr, the vessel would be held for an equivalent amount of time at constant temperature and pressure before heating or cooling is resumed at 100°F/hr. For example, the design calculations specifically considered inadvertent operation of a single bypass or solenoid-actuated relief valve leading to a 17.5°F/min blowdown for a period of 10 min to 370°F. Following this, the vessel would be held at a constant temperature of 370°F for 1 3/4 hr, then cooldown or heatup would be resumed at 100°F/hr. 1.4 Primary Coolant Leakage A double O-ring type seal is provided on the reactor vessel head closure. The area between the seals is monitored for leakage. A groove between the inner and outer O-ring communicates through the vessel flange to a line in which is installed a pressure switch between two solenoid valves. The solenoid valves are NMP Unit 1 UFSAR Section V V-6 Rev. 25, October 2017 operated from the control room. The monitoring instrumentation is shown on Figure V-1. Other primary coolant leakage is detected by monitoring leakage into the drywell floor drain tank for unidentified drywell leakage, and the drywell equipment drain tanks for identified drywell leakage. Unidentified drywell leakage from the CRDs, valve flanges, packing, component cooling water, service water, recirculation pump suction and discharge valve packing leakoff, and any other leakage not connected to the drywell equipment drain tanks, collects in the drywell floor drain tanks. Identified drywell leakage is hard piped to the drywell equipment drain tanks and includes recirculation pump seal leakage. Abnormal leakage rates for the drywell floor and equipment drain tanks are detected and alarmed in the control room. The excess leakage alarm function for the drywell floor and equipment drain tanks is performed by measuring volume changes in gallons that occur over a predetermined time period and calculating the resultant rate of change. Volume changes are used to determine the rate of change because of the irregular shape of the drywell floor and equipment drain tanks. By using volume change, excess leakage alarm capability is achieved across the entire instrument range with alarm checking occurring upon each recalculation. The rate of rise alarm function for the drywell floor drain tank is performed by measuring the amount of time between precise level step changes. When a level increase is detected, the change in tank volume and elapsed time since the last change are used to determine the rate of volume change. The rate of volume change is then used to determine the rate of rise. The calculated rate of rise is output to the control room chart recorders and alarm checked. The rate of rise for the drywell equipment drain tanks is monitored by evaluating the fill rate recorded on the equipment drain tank level chart recorder in the control room. This is performed every 4 hr. The integrated flow pumped from the drywell floor and equipment drain tanks to the waste disposal system is another means that can be used to determine leakage into the drywell floor or the equipment drain tanks. NMP Unit 1 UFSAR Section V V-7 Rev. 25, October 2017 Automatic blowdown will not occur for any primary system leak rate below the maximum allowable total operating leak rate of approximately 25 gpm. However, for breaks below about 50 gpm (although the Technical Specification limit is 25 gpm), the triple low-level setting (6 ft 3 in below minimum normal) would not be reached and automatic blowdown of relief valves would not be initiated. If normal Station offsite power were lost, both CRD hydraulic system pumps would be automatically loaded on the diesel generators to maintain water level in the vessel above the automatic blowdown trip level. It is assumed that only one CRD system is operating. The flow rate of one CRD system pump is 50 gpm at 1000 psig reactor vessel pressure and 180 gpm at zero psig reactor vessel pressure. If both pumps were operating, the flows would be greater. For much larger leak sizes, the time to reach the automatic blowdown trip level is shown in Table V-5. This table is conservatively based on only one diesel generator and its associated CRD system pump being available. 1.5 Coolant Chemistry The RCS is not designed to use inhibitors. Limits are set on chlorides, solids and gross coolant radioactivity during normal Station operation. Hydrogen water chemistry (HWC) injection and noble metal chemical addition (NMCA or NobleChem) systems are installed to reduce the potential for intergranular stress corrosion cracking (IGSCC) of the stainless steel reactor vessel components and recirculation piping. The zinc injection system is installed to reduce Cobalt 60 buildup in the primary piping corrosion films. This has the major benefit of reducing radiation dose rates in the drywell, reducing radiation exposure during outages. Hydrogen injection is provided through the feedwater/condensate systems; NobleChem is periodically added using either the classic method (injection during hot shutdown through the recirculation pump differential pressure transmitter lines) or the On-Line NobleChem (OLNC) method (with injection into feedwater during power operations); and zinc injection is provided through the feedwater system. 2.0 Reactor Vessel An isometric drawing of the reactor vessel is shown on Figure IV-9. Vessel penetrations are shown on Figure V-2 and data for the reactor vessel in Table V-1. The reactor vessel is a NMP Unit 1 UFSAR Section V V-8 Rev. 25, October 2017 vertical cylindrical pressure vessel. The base plate material is high-strength alloy carbon steel SA-302, Grade B. The vessel interior is clad with Type 308L to produce a 304 composition stainless steel following application by weld overlay. The head closure is designed for easy removal and reassembly, being bolted to the vessel with high-strength studs. Removable stud bushings are furnished in the body flange to facilitate repair of damaged threads. The CRD housings and the in-core instrumentation thimbles are welded to the bottom head of the reactor vessel. Steam outlets are from the vessel body, thus eliminating the need to break flanged joints in the steam lines when removing the vessel head for refueling. Safety valves are mounted on the vessel head. Solenoid-actuated relief valves are mounted on the main steam lines (MSL). An elevation drawing of the reactor vessel and supporting concrete structures is presented as Figure V-3. The reactor vessel is supported by a steel skirt welded to the bottom head of the vessel. The base of the skirt is continuously supported by a ring girder and sole plate fastened to a concrete foundation, which carries the load to the reactor building foundation slab. Stabilizer brackets located below the vessel flange, as shown on Figure V-3, are connected to tension bars with flexible couplings. The bars are connected to the top of the reactor shield wall. The reactor shield wall, in turn, is anchored by rigid bars to the drywell wall. The drywell wall is anchored to the concrete structure outside the drywell to limit horizontal vibration and to resist seismic and jet reaction forces. The bars and anchors to the concrete are designed to permit radial and axial movement due to thermal expansion. A head vent (Figure V-1) is provided to permit initial filling of the reactor vessel for the hydro test and to provide venting of the vessel during postaccident flooding. A drain is provided at the bottom of the reactor vessel to provide continuous flow to the cleanup system during plant operation, and to permit intermittent blowdown of crud to the drywell equipment drain tank during shutdown. This system is shown in greater detail on Figure X-2. NMP Unit 1 UFSAR Section V V-9 Rev. 25, October 2017 3.0 Reactor Recirculation Loops Five recirculation loops are provided, each of which contains a high-capacity motor-driven recirculation pump, two motor-operated gate valves for pump isolation and maintenance, and a bypass line and valve. The recirculation loops are part of the reactor primary system barrier. The recirculation pumps are single-stage, vertical, centrifugal pumps with mechanical shaft seals, driven by variable speed electric motors which receive electrical power from variable frequency motor generator (MG) sets. The pump loops are arranged within the drywell to facilitate inspection, maintenance and repair when the Station is shut down. Drywell design permits removal of the pumps and motors. Instrumentation is provided to monitor pump motor, seal and seal cooling water conditions. The valves are motor-operated gate valves with a double set of valve stem packings to provide a highly reliable seal. A valved bypass line is provided around each recirculation pump discharge valve to circulate water through the isolated loop. The large discharge valves will be opened after the water in the isolated loop reaches the bulk reactor water temperature. The piping system has support hangers which provide a constant support force on the system for all operating positions. Mechanical shock and sway suppressor devices are provided for control and protection of the system and equipment if subjected to seismic loading or vibration conditions. These devices transfer any imposed displacement forces on the piping or equipment directly to the building structure at the instant of shock occurrence, while at other times the piping or equipment is forced to move unrestricted through its normal operating range. Each recirculation loop contains coolant temperature detectors, flow monitors, differential pump pressure detectors, valve position switches, pump vibration switches, and pump motor temperature detectors. 4.0 Reactor Steam and Auxiliary Systems Piping Additional piping from the reactor vessel and recirculating system to the isolation valves is of all-welded construction and meets the Code requirements given in Table V-1. NMP Unit 1 UFSAR Section V V-10 Rev. 25, October 2017 5.0 Relief Devices Six solenoid-actuated pressure relief valves are located on the steam lines and discharge to the pressure suppression pool (Figure V-1). These valves serve two purposes: 1. Provide sufficient capacity to prevent safety valve lift for a moderate frequency event; i.e., MSIV closure with scram. 2. Depressurize the reactor vessel in the event of small system ruptures to permit timely operation of the core spray system. Three valves are required for this feature as discussed in the Loss-of-Coolant Analysis, Section XV-C. The Nuclear Regulatory Commission (NRC) has determined that safety valve discharges are not a safety concern.(1) However, the NRC would like to limit the number of discharges to the drywell. Therefore, for infrequent events, predicted peak pressures can exceed safety valve setpoints since probability of event initiation is low. For moderate-frequency events, safety valves are required not to lift. From an operational standpoint, additional pressure margins may be maintained to provide an added assurance against safety valve lift for infrequent events (safety valve lift could result in an extended outage to repair any internal containment damage versus additional margins would further reduce the probability of safety valve lift, but end-of-cycle (EOC) derates would be required, resulting in lost generation time). The 9 reactor safety valves are located on the reactor vessel head inside the primary containment (see Figure V-1). They are spring-loaded, pop-open type safety valves, which discharge directly into the containment atmosphere. The safety valves are the final protection against overpressurizing the vessel and are sized to prevent the pressure going above 1375 psig, the maximum allowed by the applicable Code (10 percent over design pressure) in the event of a MSIV closure (safety valve actuation overpressurization) event. The results of this accident are shown in the Supplemental Reload Licensing Report (SRLR)(2), and demonstrate that the peak pressure obtained in the reactor vessel is below the Code requirement of 1375 psig. NMP Unit 1 UFSAR Section V V-11 Rev. 25, October 2017 An acoustic monitoring system monitors the positions of both the relief and safety valves. This system provides an indirect indication of valve position to the control room. C. SYSTEM DESIGN EVALUATION 1.0 General Overpressure or excessive thermal cycling could lead to loss of primary system integrity. Overpressurization is prevented by a combination of automatic controls and pressure relief devices. Thermal cycling is mainly limited by operational procedures. Design limits are selected to permit possible rapid depressurization with consequent temperature transients, as might be encountered due to actuation of the solenoid relief valves. Other major considerations involve materials exposure and ability to withstand various mechanical forces. These parameters are discussed in detail in the following sections. 2.0 Pressure The reactor pressure vessel (RPV) is designed according to ASME Boiler and Pressure Vessel Code, Section I-1962, for a design pressure of 1250 psig. Pressure transients up to 10 percent over the design pressure (1375 psig) are permitted by Case 1271N applicable to this Code. The primary piping is designed and was tested according to the ASME Boiler and Pressure Vessel Code, Section I-1962, and ASA B31.1-1955, Code for Pressure Piping, for a design pressure of 1200 psig. Transients up to 15 percent over primary piping design pressures are permitted by the ASA Code for less than 10 percent of the time. Thus, short-term transients are permissible up to 1375 psig. This pressure was selected as the safety limit for the reactor primary system including both vessel and piping, even though an initial hydro test to 1875 psig was performed for the reactor vessel and 1800 psig for the reactor primary system piping. Excursions which could cause reactor overpressure are prevented by in-depth protective devices. These range from high flux scrams to utilization of the safety valves. For the most severe abnormal operational transient, the safety valve capacity is sufficient to limit the pressure to less than 110 percent of the design pressure. The devices listed in the following tabulation sequentially prevent overpressurization of the reactor vessel. Protective Devices Redundancy NMP Unit 1 UFSAR Section V V-12 Rev. 25, October 2017 High-Pressure Scram A dual fail-safe reactor protection system (RPS) with pressure transmitters in each logic channel is provided. This system is described in detail in Section VIII-A. A trip of one subchannel of two in both logic channels will result in a reactor scram; therefore, a response failure of one sensor in each logic channel would not prevent overpressure scram. Protective Devices Redundancy High Neutron Flux Also incorporated in the same manner in the RPS are high neutron flux scrams initiated by the average power range monitor (APRM) system. There are a total of eight APRM signals, four in each logic channel. The trip of only one of four in each channel will produce a scram. As discussed in Section XV, rapid operational transients of pressure and neutron flux are usually coincident. Considering the coincidence of the signals and the large amount of instrumentation provided, considerable redundancy to produce a scram exists. Solenoid-Actuated Six independently-actuated relief Relief Valves valves are provided to limit overpressure below the setpoint of the safety valves for events of moderate frequency (MSIV closure with scram), as discussed in Section XV. Safety Valves A total of 9 safety valves will limit the pressure to below 110 percent of design pressure for the MSIV closure (safety valve actuation overpressurization) event. The emergency cooling system has not been included here as a pressure-limiting device. Its capacity in terms of heat removal is only a few percent of rated capacity and cannot be utilized to reverse rapid pressure transients following isolation. The NMP Unit 1 UFSAR Section V V-13 Rev. 25, October 2017 main purpose of the emergency cooling system is to assure long-term core cooling during isolation situations by maintaining coolant inventory. 3.0 Design Heatup and Cooldown Rates The pressure vessel was fabricated in accordance with ASME Section I-1962. The nominal design temperature of the primary system corresponds to the saturation pressure of 1250 psig. Short-term temperature transients corresponding to 1375 psig, the safety limit, are also considered. Design temperature in the recirculation piping corresponds to saturated conditions at 1200 psig with short-term transients to 1375 psig also. Of all the malfunctions considered in Section XV, only the malfunction of the initial pressure regulator leads to a more rapid blowdown than the design blowdown rates of Section V-B.1.3 above. However, analyses indicate that the strains incurred are well within the 4-percent limit permitted by ASME Section III-1965 for up to ten times during the vessel lifetime. These and other extensive design analyses which determine the maximum heatup and cooldown rates are included in Section XVI. 4.0 Materials Radiation Exposure 4.1 Pressure-Temperature Limit Curves The fracture toughness requirements for the pressure vessel for testing and operational conditions are specified in Section IV of 10CFR50 Appendix G. The pressure-temperature (P-T) limit curves were developed using the methodology specified in Licensing Topical Report SIR-05-044-A (Reference 9) and ASME Code Case N-640, as well as 10CFR50 Appendix G, and the 1989 Edition of ASME Section XI, Appendix G. The bases for the technical requirements of the ASME Code are discussed in Welding Research Council (WRC) Bulletin 175. Appendix G to 10CFR50 requires that the effects of neutron irradiation on the nil ductility reference temperature (RTNDT) of the beltline materials must be included in the P-T curve calculations. Revision 2 to Regulatory Guide (RG) 1.99 is used for this purpose. Calculated adjusted reference temperature (ART) values and temperature limits are given in this section for limiting locations in the reactor vessel. Code Case N-640 permits fracture toughness curve KIc, as found in ASME Section XI, Appendix A, to be used in lieu of curve KIA of ASME Section XI, Appendix G, for development of P-T limit curves. The P-T limit curves are presented in the Pressure and Temperature Limits Report (PTLR). NMP Unit 1 UFSAR Section V V-14 Rev. 25, October 2017 The boltup limits for the flange and adjacent shell region are based on the higher of the material RTNDT of the flange and adjacent shell region (40°F) or 60°F, consistent with Reference 9. A higher minimum boltup temperature of 70°F was applied to the P-T curves, as compared to the 60°F value determined in Reference 9, in order to be consistent with the minimum bolt-up temperature value used in previous NMP1 studies. The maximum throughwall temperature gradient from continuous heating or cooling at 100°F/hr was used. The safety factors applied were as specified in the ASME Code, Section XI, Appendix G. The P-T curves are consolidated into three bounding evaluation regions of the RPV: 1) the beltline, 2) the bottom head, and 3) the feedwater nozzle/upper vessel. P-T curve calculations are performed on the beltline material which has the highest ART over the period for which the P-T curves are valid. Therefore, ART calculations were performed using RG 1.99 Revision 2 for the limiting material, plate G-8-1. Non-beltline regions are not subjected to significant fluence; therefore, RTNDT values do not change and are valid substitutions for corresponding ART values for these regions. 4.2 Temperature Limits for Boltup The reactor vessel head flange and vessel flange, in combination with the double O-ring type seal, are designed to provide a leak-tight seal when bolted together. When the vessel head is placed on the reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the O-rings of the head and vessel flange. Both the head and vessel flanges have a RTNDT of 40°F and they are not subject to any appreciable neutron radiation exposure. The minimum vessel head and head flange temperature for bolting the head flange and vessel flange is established as the higher of the material RTNDT of 40°F or 60°F, consistent with Reference 9. Consistent with other NMP1 evaluations, the minimum boltup temperature value is established at 70°F. The flanges and adjacent shell are required to be warmed to this minimum temperature of 70°F before they are stressed by the required bolt preload. A minimum temperature of 70°F is also required for the closure studs. 4.3 Temperature Limits for In-Service System Pressure Tests NMP Unit 1 UFSAR Section V V-15 Rev. 25, October 2017 The fracture toughness analysis for in-service system pressure tests with fuel in the vessel resulted in a revised set of P-T limits shown in the PTLR. The calculated adjustment to the RTNDT, based on Revision 2 of RG 1.99, is used in the analysis to account for the effect of fast neutrons. 4.4 Operating Limits During Heatup, Cooldown, and Core Operation The fracture toughness analysis was done for the assumed heatup or cooldown rate of 100°F/hr. The temperature gradients and thermal stress effects corresponding to this rate were included. In order to assess the ART at the 1/4T and 3/4T positions, RG 1.99 requires an assessment of the peak fast (E>1Mev) neutron flux at the inner diameter (ID) surface of the pressure vessel. These data were determined using RG 1.190 compliant plant-specific methods benchmarked to Unit 1 flux monitors that had been removed and tested. 4.5 Predicted Shift in RTNDT The allowable internal vessel pressure for a specific coolant temperature is a function of several key variables including the ART. The ART for the vessel beltline region enters the P-T calculations directly via the reference fracture toughness curve (KIc). Therefore, it is necessary to provide reasonable and conservative estimates of the shift in RTNDT for the period of time for which the P-T calculations will be used. The ART was calculated using Revision 2 to RG 1.99. The vessel material surveillance program is outlined in Section XVI. 4.6 Neutron Fluence Calculations Reactor vessel neutron fluence has been evaluated using a method in accordance with the recommendations of RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2001. Future evaluations of reactor vessel fluence will be completed using a method in accordance with the recommendations of RG 1.190 (as noted in Reference 5). NRC approval of the Unit 1 neutron fluence calculational methodology is documented in Reference 6. 5.0 Mechanical Considerations NMP Unit 1 UFSAR Section V V-16 Rev. 25, October 2017 5.1 Jet Reaction Forces The RPV and support structures are designed to withstand the forces that would be created by full area flow of any vessel nozzle with the RPV at design pressure. Thus, even if one line ruptured, the vessel would not be moved by jet reaction forces sufficiently to cause rupture of other connected pipes. 5.2 Seismic Forces The reactor primary system is designed and constructed in accordance with performance objective 1 for seismic design, as described in Section III. 6.0 Safety Limits, Limiting Safety Settings and Minimum Conditions for Operation Safety limits are appropriate for maximum pressure and heatup and cooldown rates for the RCS. The pressure is automatically limited by devices for which limiting safety system settings are required. These settings include safety valve actuation, reactor high-pressure scram, and reactor high-flux scram. Minimum conditions for operation are appropriate for both the devices which have limiting safety system settings and for other pertinent operating parameters. These minimum conditions for operation include the number of operable reactor safety valves, the number of operable solenoid-actuated relief valves, maximum permissible values for various coolant chemical parameters and allowable heatup and cooldown rates. D. TESTS AND INSPECTIONS 1.0 Prestartup Testing The RCS was given a system hydrostatic test in accordance with Code requirements prior to initial reactor startup. Before pressurization, the system was heated to RTNDT + 60°F, i.e., 100°F. Piping and support hangers were checked while thermal expansion was in progress. Further details of initial tests are given in Section XIV. 2.0 Inspection and Testing Following Startup Tests and inspections of the safety and relief valves are covered by the Inservice Inspection (ISI) and Inservice Testing (IST) Programs. Reactor pressure and flux scram components are NMP Unit 1 UFSAR Section V V-17 Rev. 25, October 2017 tested as per the Technical Specifications. Visual inspections are also performed as covered in the ISI Program. In addition, the following tests and inspections are performed. 2.1 Pressure Test A leakage test at operating pressure is made on the primary system following each removal and replacement of the reactor vessel head. The system is checked for leaks and abnormal conditions are corrected before reactor startup. The minimum vessel temperature during the leakage test is at least 70°F prior to pressurizing the vessel. 2.2 Pressure Vessel Irradiation Vessel material surveillance samples are located within the reactor vessel to enable periodic monitoring of material properties with exposure. The vessel material surveillance program is outlined in Section XVI. 2.3 Reactor Coolant System Pressure Isolation Valve Leakage Testing The pressure isolation valves (PIV) function is to prevent intersystem overpressurization between the RCS and connected low pressure systems. The PIVs, listed in Table V-6, are included in the IST program and are required to be leak tested in accordance with the Technical Specifications. The list of PIVs in Table V-6 was initially identified in Reference 7, and was subsequently supplemented by Reference 8. E. EMERGENCY COOLING SYSTEM 1.0 Design Bases The emergency cooling system provides for decay heat removal from the reactor fuel in the event that reactor feedwater capability is lost and the main condenser is not available. Performance criteria 2 and 6 (V-A.2.0) are met by inclusion of this system. Approximately 48 hr of makeup water is available to the emergency condensers (EC) from the condensate storage tanks (CST). One EC system (i.e., two condensers) has a heat removal capacity at normal pressure of 19.0 x 107 Btu/hr, which is approximately 3 percent of maximum reactor thermal power. This capacity is sufficient to handle the decay heat production at 100 sec following a scram. Either half of the emergency cooling system may be independently isolated. Pertinent design NMP Unit 1 UFSAR Section V V-18 Rev. 25, October 2017 information is included in Table V-1. Decay heat loads were calculated from the work of Shure(3,4), corrected for U-239 and Np-239. 2.0 System Design and Operation The emergency cooling system is connected to the reactor and operates by natural circulation. It serves as an alternate heat sink when the reactor is isolated from its normal heat sink (the main condenser). A flow diagram is included on Figure V-1. Each of the two independent emergency cooling loops includes two condensers consisting of a tube bundle in a tank located at floor el 340', which is above the reactor vessel. A minimum of 10,680 gal is maintained in each tank. A maximum EC shell side level is equal to the elevation of the shell overflow lines. Each loop has an elevated 40,000-gal makeup water storage tank with gravity feed to the condensers. Both makeup tanks in the system are cross-tied via a common line, which allows the system to operate using only one loop while utilizing both makeup tanks. During operation of the emergency cooling loops, steam rises from the reactor vessel to the condenser tubes where it is condensed by boiling the condenser shell water at approximately 5 psig. As the water condenses, it returns by gravity flow to the suction of a reactor recirculating pump and, thus, to the reactor vessel. In the standby condition, the steam inlet isolation valves are normally open so that the tube bundles are continuously at reactor pressure. The condensate outlet isolation valves are closed so as to maintain the tubes in each EC flooded on the tube side. To compensate for outlet isolation valve leakage, a keep-full system has been added to ensure that the tube side of the EC remains flooded in the standby condition. To maintain adequate keep-full flow, the existing temperature switches and sensors on the EC steam inlet lines will be used to trigger alarms on high and low level conditions. New thermocouples were added to monitor water level in the EC steam supply line to ensure the condenser tubes remain covered, and to monitor spillover into the horizontal leg to evaluate any potential thermal stresses. By maintaining tube side water level above the tube bundle in the standby condition, the amount of thermal energy wasted and thermal stresses on the tubes are greatly reduced. The system is placed into operation by opening the normally closed condensate return isolation valve, which is dc NMP Unit 1 UFSAR Section V V-19 Rev. 25, October 2017 solenoid air operated and will fail open in the event of a total loss of dc power or air pressure, thus putting the system in operation. Automatic operation of the emergency cooling system is initiated by high reactor pressure in excess of 1080 psig, sustained for 12 sec. The time delay is provided to prevent unnecessary actuation of the system during anticipated turbine trips. To assist in depressurization for small breaks, the system is initiated on low-low reactor water level, 5 ft below minimum normal water level (5-in indicated scale), sustained for 12 sec. The system may also be initiated manually, either from the main control room or independently from two remote shutdown panels (RSP) (one panel for each emergency cooling loop). The RSPs are described in Section X-L. During operation, water on the shell side of the condensers boils and vents to atmosphere while condensing steam inside the tube bundles. Radiation monitors are located on the vent to detect tube bundle leaks during system operation from the offending half of the system. Level control valves allow makeup water to drain from the elevated 40,000-gal makeup storage tanks to the condensers to maintain the 6-in level tolerance. The condensers, combined with both makeup tanks, can provide continuous cooling for 8 hr. Normally, water will be supplied to the tanks through the condensate transfer system from the two 200,000-gal CSTs. Thus, approximately 48 hr of continuous cooling is possible. Overflow from the shell side of the condensers is drained to the waste collector tank. The electric and diesel-driven fire pumps are also available to supply the makeup tanks and, thus, the condenser shells. Vents are provided at the high points in the steam lines to purge noncondensable gases from the reactor vessel which may inhibit core cooling during natural circulation. One vent connects to the MSL beyond the second isolation valve and is used during normal operation. A second vent is provided to the pressure suppression chamber which would be used during accident conditions to vent the reactor vessel if needed. Drains are provided at the low points in the steam lines to eliminate condensate from the system which may cause flashing/water hammer at low reactor pressures. 3.0 Design Evaluation 3.1 Redundancy NMP Unit 1 UFSAR Section V V-20 Rev. 25, October 2017 One-half of the emergency cooling system is adequate to handle the decay heat following reactor isolation from infinite operation at 1850 MWt. A rapid isolation due to MSIV closure or turbine trip without bypass will open the solenoid-actuated relief valves for a short time (10-15 sec) to reverse the pressure transient. Coolant loss during this time is less than 1000 lb. If only one loop is operational, 2000 lb of coolant will be lost through the relief valves during the first 100 sec until decay heat can be handled by one loop. This loss of coolant is relatively small compared to the total primary system inventory of 450,000 lb. Subsequent to this transient, and even in the event of total ac power loss with no makeup whatsoever to the reactor vessel, operation of the emergency cooling system will proceed normally. 3.2 Makeup Water Sufficient makeup water is available from the gravity feed tanks and the CSTs, via the condensate transfer pumps, to permit operation for 48 hr. These pumps can be powered from the Station diesel generators. An alternate source of water to the makeup tanks is also available from the electric and diesel fire pumps. 3.3 System Leaks In the event of a tube failure in the EC during system operation, radiation monitors on the vent lines to atmosphere provide alarm indication in the control room. Closure of the inlet and outlet valves to the offending half of the system is initiated manually. A flow detector is provided in the inlet line to each half of the system. The detector initiates closure of isolation valves in the affected condenser loop in the event of a line break, resulting in a flow of about 300 percent of rated flow. This is equivalent to a rupture of 30 percent of the cross-sectional area of the pipe. Local temperature detectors (at the isolation valves) are provided to detect small leaks and provide alarm functions. The common vent line to the MSL is not affected. For the steam supply and condensate return isolation valves, additional signals are required from confirmatory logic relays in the reactor building to prevent inadvertent isolation of the EC loops. For example, a spurious isolation signal might have resulted from a fire in the control complex. NMP Unit 1 UFSAR Section V V-21 Rev. 25, October 2017 3.4 Containment Isolation In each of the two emergency cooling system steam supply lines, both isolation valves are located outside the drywell, with the body of the valve closest to the drywell shell welded directly to the penetration sleeve by a full-penetration weld (Figure V-8). The valve body forms a part of the containment wall. For the recirculation system line break, containment can be ensured by remote manual isolation of these lines should the integrity of the emergency cooling system pressure boundary be compromised. If an emergency cooling system line were to break inside the containment, the isolation valves would prevent fission product escape through the line but would not prevent loss of fluid or fission products to the containment. If the line were to break outside the containment, the isolation valves would limit loss of fluid and fission products from the reactor. The only malfunction which could result in an uncontrolled release of coolant and fission products outside the containment would be a rupture of the body of the valve adjacent to the drywell shell. This failure is considered incredible due to the stringent design, fabrication, and inspection techniques utilized. 4.0 Tests and Inspections 4.1 Prestartup Test A heat removal capacity test of the emergency cooling system was made during initial startup. A subsequent capacity test was conducted upon startup following replacement of the EC tube bundles (Forced Outage 97-07). 4.2 Subsequent Inspections and Tests Surveillance requirements, including tests for operability of isolation valves and condensate transfer pumps, and calibration of level, pressure, and radiation instrumentation, are performed at frequent intervals to assure the reliability of the system when operation is required. The system heat removal capability is determined at 5-yr intervals. Test valves installed downstream of each isolation valve permit these isolation valves to be leak tested on a regular basis. NMP Unit 1 UFSAR Section V V-22 Rev. 25, October 2017 The pressure testing requirements for the ECs will meet the ASME Code requirements stated in the Nine Mile Point Nuclear Station -Unit 1 (Unit 1) Pressure Testing Program Plan. During hydrostatic testing with the reactor not critical and reactor coolant temperature greater than 212°F, the emergency cooling system is not required. F. REFERENCES 1. NRC Standard Review Plan, Section 5.2.2, Overpressurization Protection - NUREG 75-087. 2. "Supplemental Reload Licensing Report for Nine Mile Point 1 Reload 24 Cycle 25," 002N6949, Revision 0, March 2017. 3. "Calculating Energy Release by Fission Products," AEC Report WAPD-T-1309, March 1961. 4. K. Shure, "Fission Product Decay Heat," AEC Report WAPD-BT-24, December 1961. 5. NRC Letter to NMPNS dated November 8, 2004, "Nine Mile Point Nuclear Station Unit Nos. 1 and 2 - Issuance of Amendments RE: Implementation of the Reactor Pressure Vessel Integrated Surveillance Program (TAC Nos. MC1758 and MC1759)." 6. NRC Letter to NMPNS dated October 27, 2003, "Nine Mile Point Nuclear Station, Unit No. 1 - Issuance of Amendment Re: Pressure-Temperature Limit Curves and Tables (TAC No. MB6687)." 7. NRC Letter to NMPC dated April 20, 1981, "Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves." 8. NRC Letter to NMPC dated March 20, 1995, "Issuance of Amendment for Nine Mile Point Nuclear Station Unit No. 1 (TAC No. M89786)." 9. Structural Integrity Associates Report No. SIR-05-044-A, Revision 0, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," April 2007. NMP Unit 1 UFSAR Section V V-23 Rev. 25, October 2017 TABLE V-1 REACTOR COOLANT SYSTEM DATA Reactor Vessel Internal height 63 ft 10 in Internal diameter 17 ft 9 in Design pressure and temperature 1250 psig at 575°F Operational heatup and cooldown 100°F/hr rates Base metal material SA 302B (modified) Wall base metal thickness 7 1/8 in Top head base metal thickness 4 5/16 in (minimum) Bottom head base metal thickness 8 3/4 in (minimum) Vessel design lifetime 40 years Estimated lifetime neutron 6 x 1017nvt* fluence (energies >1 Mev) Initial RTNDT (base metal 36°F (limiting plate opposite core) G-8-1) ART at 36/46 EFPY 161.6°F/171.2°F (limiting plate G-8-1) Cladding material Weld deposited 308L electrode Cladding thickness, nominal 7/32 in Nozzle material ASTM SA 336 modified and SA 336 F8 and F8M Vessel head stud material ASME SA 540 Grade B23 or B24 Class 3 Design codes ASME Sec I-1962 and Cases 1270N & 1273N Recirculation Loops Number 5 Material TP 316NG Design code ASME Sec III-1977 through Winter 1979 addenda Design pressure and temperature 1200 psig at 570°F Recirculation Pumps Number 5 Type Vertical centrifugal Power rating 1000 hp Speed 890 rpm Flow rate 7,200 to 36,000 gpm Water temperature (normal) 530°F Design pressure and temperature 1300 psig at 575°F Required inlet head 48 ft water Outlet head 89 to 120 ft water NMP Unit 1 UFSAR Section V V-24 Rev. 25, October 2017 TABLE V-1 (Cont'd.) Recirculation Pumps (cont'd.) Casing material SA 351 CF-8M Casing design pressure 1300 psig Design code ASME Sec I-1962, Standard of the Hydraulic Institute, ASA B31.1-1955 Recirculation Loop Valves Number 3 per loop (1 bypass) Type Gate Casing material SA 351 CF-8M Design code ASME Sec I-1962 and ASA B31.1-1955 Steam Lines Number 2 Material ASME A106 Grade B Design code (to outer ASME Sec I-1962 and ASA isolation valve) B31.1-1955 with the requirements of ASME Sec III-1965 for nondestructive testing Steam Line Isolation Valves Number 2 per line Type 1 air-operated wye globe (outside of drywell) 1 ac motor-operated globe (inside of drywell) Casing material A216 Grade WCB Closing time: air-operated 3 - 10 sec (adjustable) motor-operated 10 sec Design code ASME Sec I-1962 Relief Valves Number 6 Capacity (each, nameplate) 540,910 lb/hr @ 1120 psig (each, analytical 540,900 lb/hr @ 1120 psig maximum) Pressure setting (rated power) 2 @ 1090 psig 2 @ 1095 psig 2 @ 1100 psig Design code ASME Sec III-1968 NMP Unit 1 UFSAR Section V V-25 Rev. 25, October 2017 TABLE V-1 (Cont'd.) Safety Valves Number 9 Capacity (each) 633,000 to 651,000 lb/hr Pressure setting (nominal) 1218 to 1254 psig Capacity (minimum each, 644,543 lb/hr at 1278 psig certified) Design code ASME Sec I-1962 Emergency Cooling System Condensers: Design pressure - shell 15 psig at 300°F - tubing 1250 psig at 575°F Design code - shell ASME Sec VIII (nuclear cases) and ASME Sec III, Subsection ND, 1986 edition - tubing ASME Sec III, Subsection NC, Class 2, 1986 edition Number of tube bundles 4 Capacity (rated capacity of 38 x 107 Btu/hr at 1135 four units) psig and 562°F on tube side; 5 psig and 228°F on shell side Operating time with gravity 8 hr makeup Isolation valves in inlet line 2 normally open motor operated Isolation valves in outlet line 1 normally closed air operated and 1 check valve System Pressures Design 1250 psig at 575°F Initial vessel hydrostatic test 1875 psig pressure Maximum safety valve setting 1254 psig Minimum safety valve setting 1218 psig Solenoid-actuated relief valve 2 @ 1090 psig settings 2 @ 1095 psig 2 @ 1100 psig Emergency cooling system 1080 psig for 12 sec pressure actuation Reactor scram 1080 psig Normal operating pressure 1030 psig at 550°F

• This value represents the original design estimate of lifetime neutron fluence for the reactor vessel.

NMP Unit 1 UFSAR Section V V-26 Rev. 25, October 2017 TABLE V-2 OPERATING CYCLES AND TRANSIENT ANALYSIS RESULTS** OPERATING CYCLE EXPECTED NO. OF CYCLES Vessel Head Removal 50 Vessel Head Reinstallation 50 100°F/hr Heatup 240 100°F/hr Cooldown* 229 300°F/hr Emergency Cooldown 10 Blowdown 1 Scram Cycles 280 Emergency Condenser Initiation into Isolated Loop 30 Unisolation of an Isolated Loop 30 Emergency Condenser Initiation into Idle Loop 30 Shutdown Cooling Initiation into Isolated Loop 240 Inadvertent Start of Cold Loop 20 Emergency Condenser into Pumped Loop 500 Recirculation Pump Hot Loop Startup 300

• The number of 100°F/hr cooldowns was determined by subtracting the emergency cooldowns and blowdown from the number of 100°F/hr heatups.

NMP Unit 1 UFSAR Section V V-27 Rev. 25, October 2017 TABLE V- ** This table was used in the original fatigue evaluations of the reactor vessel as detailed in Section V-A.4.0. The NMP Fatigue Monitoring Program is now used to manage the fatigue evaluations at NMP. This program uses the FatiguePro software to maintain all actual calculated cumulative usage factors below their corresponding allowables. The calculations are based on the actual number of cycles experienced by the plant for each transient and, in some cases, the actual parameters experienced by each cycle rather than the number of cycles listed in this table. NMP Unit 1 UFSAR Section V V-28 Rev. 25, October 2017 TABLE V-3 FATIGUE RESISTANCE ANALYSIS Region of Vessel Usage Factor* Closure Studs 0.205 Basin Seal Skirt Weld 0.782 Feedwater Nozzles With Repair Cavities 0.489 Without Repair Cavities 0.163 Control Rod Drive Penetrations 0.060 Lower Vessel Head, Vessel Support 0.0833 Skirt and Core Support Cone Reactor Recirculation Nozzles 0.006

• Listed usage factor values are based on original fatigue evaluations of the reactor vessel as detailed in Section V-A.4.0. Values are managed and maintained below their corresponding allowables through the NMP Fatigue Monitoring Program.

NMP Unit 1 UFSAR Section V V-29 Rev. 25, October 2017 TABLE V-4 CODES FOR SYSTEMS FROM REACTOR VESSEL CONNECTION TO SECOND ISOLATION VALVE Piping Vessel Nozzle to Second Isolation Valve Isolation Valves Shutdown Cooling ASA B31.1-1955; ASME Sec I-1962 Cleanup ASME Sec I-1962 and Articles N324 and N460 to N469 of ASME Sec III-1965; ASME Sec III, Appendix F, 1986 Edition* ASA B31.1-1955; ASA B31.1-1955, ASME Sec I-1962 and certain requirements Articles N324 and of ASME Sec N460 to N469 of IIIA-1965, and ASME ASME Sec III-1965; Sec III-1986 ASME Sec III, (IV-38-13) Appendix F, 1986 Edition* Feedwater ASA B31.1-1955; ASME Sec I-1962 Core Spray ASME Sec I-1962 and Articles N324 and N460 to N469 of ASME Sec III-1965 ASA B31.1-1955; ASA B31.1-1955 and ASME Sec I-1962 and certain requirements Articles N324 and of ASME Sec IIIA-1965 N460 to N469 of ASME Sec III-1965; ASME Sec III, Appendix F, 1986 Edition* Liquid Poison ASA B31.1-1955, ASME Sec I-1962 ASME Sec I-1962 and Articles N324 and N460 to N469 of ASME Sec III-1965

• For analyzing thermally-induced overpressurization conditions between isolation valves.

NMP Unit 1 UFSAR Section V V-30 Rev. 25, October 2017 TABLE V-5 TIME TO AUTOMATIC BLOWDOWN Total Primary Time to Reach System Leak Rate Automatic Blowdown (gpm) Trip Level 80 36 min 135 21 min 225 14 min 510 7 min 2,835 75 sec 5,685 40 sec NMP Unit 1 UFSAR Section V V-31 Rev. 25, October 2017 TABLE V-6 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES System Valve No. Core Spray 40-03 40-13 Condensate Supply to 40-20 Core Spray (Keep Fill 40-21 System) 40-22 40-23 Core Spray Supply to 38-165 Shutdown Cooling 38-166 (Water Seal) 38-167 38-168 38-169 38-170 38-171 38-172 REACTOR EMERGENCY COOLANT SYSTEM r-M[NSTEllH llflA!NLllE i ; ' ' El llfT£RCONIECTIONTODl.PLICATE SET OF EMEllOENCY CCIO!NSl:RS TO MAIN STEAM V£N1LlNE 8-r---r----, T ... 1 e _..._.,._J e ' ' ' ' ' ' t HMEl.l'Flltfol TU'llllNE 9UILOl"-l HAIN STElllM L!HES TOTLlllllNE SE£F[(i,Xl-i FIGURE V-1 UFSAR Rev1s1on 19 October 2005 NOZZLE NUMBER NIA THRU NIE N2A THRU N2E N3A,N3B N4A THRU N40 N5A,N5B5 N6A,N68 N7A THRU N7U NB N'l N10 Nil N12 N13A Nl3B NJ4A N!48 Nl5A NJ58 NJ6A Nl6B N17A Nl7B NIB QTY. SIZE 5 36' TO 28' 5 28' 2 24' SPCL. 4 10' SPCL. 2 10' 2 6' 18 6' 1 4' 1 3' SPCL. 129 6' SPCL. 64 2' SPCL. I 1 I 1 1 1 I I I I 1 I I 1 REACTOR VESSEL NOZZLE LOCATION FUNCTION AZIMUTH ELEV. RECIRCULATION OUTLET 42° .114° ,186° ,258° .330° 14'-0' RECIRCULATION INLET 0° ,72° .144° ,216° ,288° STEAM OUTLET -TURBINE 90° .270° 49'-3' FEEDWATER 45° ,135° ,225° ,315° 35'-2' STEAM OUTLET-70°,290° 45'-1' EMER.CONOENSER CORE SPRAY 60° ,240° 34'-0' 9 SAFETY VALVES, I CWIOE RANGE LEVEL-TOP) ANO 8 BLIND FLANGES VENT CONTROL ROD DRIVE-270° 35'-2' HYDR.SYSTEM RETURN CONTROL ROD DRIVE FLUX MONITOR HIGH PRESSURE SEAL 00 LEAK DETECTOR LOW PRESSURE SEAL 5°-37W LEAK DETECTOR CORE DIFFERENTIAL PRESS. 98° PROTECTION SYSTEM REF. 57° 46'-l' COLUMN-TOP PROTECTION SYSTEM REF. 62° 35'-2' COLUMN-BOTTOM LEVEL CONTROL REF. POT. B4° 45'-l' LEVEL CONTROL STATIC B2° 35'-2' LEG LEVEL CONTROL REF.POT. 236° 45'-l' LEVEL CONTROL STATIC 236° 35'-2' LEG PROTECTION SYSTEM REF. 244° 46'-l' COLUMN-TOP PROTECTION SYSTEM REF. 244° 35'-2' COLUMN-BOTTOM INACTIVE 50'-5' WIOE RANGE LEVEL-252° 35'-2' BOTTOM DRAIN 315° IUTIHC SURF AC! EUY. -II H 32" "9.IU L 8 JU" EMA. HOLES fQUAl.L Y SPACED STRADDLE o0' un° T TOP OF ACTIV[ rn£L ELEV. 29' -4 1118 .. 1--------Jr .g* I BOTTOM CF A.CTl'YE FUEL ELEV. 11' -5 5/16" FIGURE V-2 UFSAR Rev1s1on 16 November 1999

Nine Mile Point Unit 1 UFSAR FIGURE V-4 THRU FIGURE V-7 FIGURES V-4 THRU V-7 HAVE BEEN DELETED UFSAR Revision 16 1 of 1 November 1999 EMERGENCY CONDENSER SUPPLY ISOLATION VALVES <TYPICAL OF 2> GUARD PIPE----. NOZZLE---. FIGURE V-8 UFSAR REVISION 16 NOVEMBER 1qqq

NMP Unit 1 UFSAR Section VI EF VI-1 Rev. 25, October 2017 SECTION VI LIST OF EFFECTIVE FIGURES Figure Number Revision Number VI-1 14 VI-2 14 VI-3 14 VI-4 14 VI-4a 14 VI-5 14 VI-6 14 VI-7 14 VI-8 14 VI-9 14 VI-10 14 VI-11 14 VI-12 14 VI-13 14 VI-14 14 VI-15 14 VI-16 14 VI-17 14 VI-18 14 VI-19 14 VI-20 17 VI-21 14 VI-22 17 VI-23 16 VI-24 16 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section VI VI-i Rev. 25, October 2017 SECTION VI CONTAINMENT SYSTEM A. PRIMARY CONTAINMENT - MARK I CONTAINMENT PROGRAM 1.0 General Structure 2.0 Pressure Suppression Hydrodynamic Loads 2.1 Safety/Relief Valve Discharge 2.2 Loss-of-Coolant Accident 2.3 Summary of Loading Phenomena 3.0 Plant-Unique Modifications B. PRIMARY CONTAINMENT - PRESSURE SUPPRESSION SYSTEM 1.0 Design Bases 1.1 General 1.2 Design Basis Accident (DBA) 1.3 Containment Heat Removal 1.4 Isolation Criteria 1.5 Vacuum Relief Criteria 1.6 Flooding Criteria 1.7 Shielding 2.0 Structure Design 2.1 General 2.2 Penetrations and Access Openings 2.3 Jet and Missile Protection 2.4 Materials 2.5 Shielding 2.6 Vacuum Relief 2.7 Containment Flooding C. SECONDARY CONTAINMENT - REACTOR BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings 1.2 Pressure Relief Design 1.3 Seismic Design 1.4 Shielding 2.0 Structure Design 2.1 General Structural Features D. CONTAINMENT ISOLATION SYSTEM 1.0 Design Bases 1.1 Containment Spray Appendix J Water NMP Unit 1 UFSAR Section Title Section VI VI-ii Rev. 25, October 2017 Seal Requirements 2.0 System Design 3.0 Tests and Inspections E. CONTAINMENT VENTILATION SYSTEM 1.0 Primary Containment 1.1 Design Bases 1.2 System Design 2.0 Secondary Containment 2.1 Design Bases 2.2 System Design F. TEST AND INSPECTIONS 1.0 Drywell and Suppression Chamber 1.1 Preoperational Testing 1.2 Postoperational Testing 2.0 Containment Penetrations and Isolation Valves 2.1 Penetration and Valve Leakage 2.2 Valve Operability Test 3.0 Containment Ventilation System 4.0 Other Containment Tests 5.0 Reactor Building 5.1 Reactor Building Normal Ventilation System 5.2 Reactor Building Isolation Valves 5.3 Emergency Ventilation System G. REFERENCES NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section VI VI-iii Rev. 25, October 2017 VI-1 DRYWELL PENETRATIONS VI-2 SUPPRESSION CHAMBER PENETRATIONS VI-3a REACTOR COOLANT SYSTEM ISOLATION VALVES VI-3b PRIMARY CONTAINMENT ISOLATION VALVES - LINES ENTERING FREE SPACE OF THE CONTAINMENT VI-4 SEISMIC DESIGN CRITERIA FOR ISOLATION VALVES VI-5 INITIAL TESTS PRIOR TO STATION OPERATION NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section VI VI-iv Rev. 25, October 2017 VI-1 DRYWELL AND SUPPRESSION CHAMBER VI-2 ELECTRICAL PENETRATIONS - HIGH VOLTAGE VI-3 ELECTRICAL PENETRATIONS - LOW VOLTAGE VI-4 PIPE PENETRATIONS - HOT VI-4a CLAMSHELL EXPANSION JOINT VI-5 TYPICAL PENETRATION FOR INSTRUMENT LINES VI-6 REACTOR BUILDING DYNAMIC ANALYSIS - ACCELERATION EAST-WEST DIRECTION VI-7 REACTOR BUILDING DYNAMIC ANALYSIS - DEFLECTIONS EAST-WEST DIRECTION VI-8 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING SHEAR EAST-WEST DIRECTION VI-9 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING MOMENT EAST-WEST DIRECTION VI-10 REACTOR BUILDING DYNAMIC ANALYSIS - ACCELERATION NORTH-SOUTH DIRECTION VI-11 REACTOR BUILDING DYNAMIC ANALYSIS - DEFLECTIONS NORTH-SOUTH DIRECTION VI-12 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING SHEAR - NORTH-SOUTH DIRECTION VI-13 REACTOR BUILDING DYNAMIC ANALYSIS - ELEVATION VS. BUILDING MOMENT - NORTH-SOUTH DIRECTION VI-14 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. ACCELERATION VI-15 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. DEFLECTION NMP Unit 1 UFSAR Figure Number Title Section VI VI-v Rev. 25, October 2017 VI-16 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. SHEAR VI-17 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. MOMENT VI-18 TYPICAL DOOR SEALS VI-19 DETAILS OF REACTOR BUILDING AIR LOCKS VI-20 INSTRUMENT LINE ISOLATION VALVE ARRANGEMENT VI-21 TYPICAL FLOW CHECK VALVE VI-22 ISOLATION VALVE SYSTEM VI-23 DRYWELL COOLING SYSTEM VI-24 REACTOR BUILDING VENTILATION SYSTEM NMP Unit 1 UFSAR Section VI VI-1 Rev. 25, October 2017 SECTION VI CONTAINMENT SYSTEM A pressure suppression containment system consisting of a drywell, suppression chamber (torus), and interconnecting vent piping is the primary containment for the main coolant system. When the reactor is hot (>215°F) and pressurized, the reactor building containing the pressure suppression system provides a secondary containment barrier. When the reactor is shut down for refueling, maintenance, or testing, and the drywell head is removed, or pressure suppression system integrity is not required, the reactor building provides the primary containment. A. PRIMARY CONTAINMENT - MARK I CONTAINMENT PROGRAM 1.0 General Structure The objective of the Mark I Containment Program is to demonstrate that all Mark I containments have acceptable structural margins throughout their design life when compared to criteria acceptable to the Nuclear Regulatory Commission (NRC). This program consists of testing and analysis of both structural and hydrodynamic phenomena; also, it addresses the effect of structural/hydrodynamic phenomena on containment loads. It includes the establishment of structural acceptance criteria against which the results of structural evaluations can be assessed. The program includes an evaluation of the need for structural modifications and/or load mitigation devices to assure adequate structural margins. Key elements of the program are: Load Definition Report - Documentation of the design basis hydrodynamic pressure suppression loads and their possible combinations. Structural Acceptance Criteria - Identification of the acceptance criteria against which the structural evaluation results will be assessed. They will consider current requirements and increased knowledge gained since original design, including specific test support as required. Plant-Unique Analyses - Specific structural evaluation of each plant by using the loads defined in the Load Definition Report in conjunction with the established Acceptance Criteria. NMP Unit 1 UFSAR Section VI VI-2 Rev. 25, October 2017 The plant-unique analyses reports will be submitted by each utility to the NRC for review and approval. This approval of plant-unique analysis reports with any required structural modifications and/or load mitigation devices and safety evaluation reports completes the program. 2.0 Pressure Suppression Hydrodynamic Loads Hydrodynamic loads to which the pressure suppression system can be subjected are due primarily to the following phenomena: 1) safety/relief valve (SRV) discharge, and 2) loss-of-coolant accident (LOCA). 2.1 Safety/Relief Valve Discharge Actuation of a SRV produces a dynamic loading on components and structures in the suppression pool region. When a relief valve lifts, the effluent reactor steam causes a rapid pressure buildup in the discharge pipe, due to compression of the column of air initially occupying the pipe, and a subsequent acceleration of the water slug in the submerged portion of the pipe. During this process, the pressure in the pipe builds to a peak as the last of the water is expelled. At this point, the compressed air between the water slug and the effluent vapor begins to leave the pipe. As the compressed air exits the discharge line it immediately begins to expand, displacing the water and propagating a pressure disturbance throughout the suppression pool. The dynamics of expanding a compressed air bubble result in pressure oscillations (similar to that of a spring-mass system) arising from the bubble expansion coupled with inertial effects of the moving water mass. The magnitude of the pressure disturbance in the suppression pool decreases with increasing distance from the point of discharge, resulting in a damped oscillatory load at every point on the torus wall below the water surface. This load produces oscillatory stresses in the torus shell. There are several SRVs in the plant, each having different discharge line characteristics, but the above general description is applicable. Additional types of actuation to be considered are: Consecutive actuation (one valve actuating several times). Multiple actuation (two or more valves actuating simultaneously). NMP Unit 1 UFSAR Section VI VI-3 Rev. 25, October 2017 Multiple consecutive actuation (two or more valves actuating several times). 2.2 Loss-of-Coolant Accident The various phenomena that can occur during the course of a postulated LOCA in a Mark I pressure suppression containment system can result in dynamic loads on the torus and its associated structures. With a postulated instantaneous rupture of a steam or recirculation line, the escaping steam-water mixture would cause a very rapid increase in drywell pressure and temperature. As the drywell pressure increases, the water initially in each downcomer accelerates into the pool and each downcomer clears of water. During this water clearing process, a jet can form in the suppression pool which may cause water jet impingement loads on the structures within the suppression pool and on the torus. Immediately following downcomer clearing, a bubble of air from the drywell starts to form at the exit of the downcomers. Since initially the bubble pressure is essentially equal to the drywell pressure at the time of clearing, the bubble pressure is transmitted through the suppression pool water and results in a downward load on the torus. When the air-steam mixture flows from the drywell through the vent system, the bubble initially formed expands and decompresses. Continued injection of drywell air and expansion of the air bubble result in a rise of the suppression pool surface. Structures close to the pool surface experience loads as the rising pool surface impacts the lower surface of the structure. As the suppression pool surface rises, the air in the upper half of the torus is compressed and causes a net upward load on the torus. As the pool surface rises, the air bubble passes through the water ligament and there is a breakup of the water slug. The subsequent pool swell evolves into a two-phase "froth" of air and water. The pool swell transient associated with drywell air venting to the pool typically lasts for 3 to 5 sec. Following air carryover, there will be a period of decreasing steam flow rate through the vent system. This time period has been subdivided into three phases: 1) high mass flux, characterized by nearly steady-state condensation; 2) medium mass flux, characterized by periodic variations in condensation rate; and 3) low mass flux chugging, characterized by intermittent condensation. NMP Unit 1 UFSAR Section VI VI-4 Rev. 25, October 2017 During steam condensation the downcomers experience a lateral loading caused by random movement of the steam-water interface. The magnitude of this load varies with steam mass flux and suppression pool temperature. The maximum lateral loads in a design basis LOCA will occur toward the end of blowdown. Shortly after a LOCA, the emergency core cooling system (ECCS) pumps have automatically started to pump suppression pool water into the reactor pressure vessel (RPV). This water drains through the core and into the drywell from the break. Because the drywell will be full of steam when the vessel floods, the introduction of water causes steam condensation and drywell depressurization. 2.3 Summary of Loading Phenomena The following is a listing of the various loads which may be experienced by the containment system due to SRV discharge and LOCA phenomena: Safety/Relief Valve Water clearing loads Air clearing loads Steam flow condensation loads Submerged structure loads - velocity and acceleration drag loads Thrust loads on safety/relief valve discharge lines (SRVDL) SRVDL internal pressure Pool stratification effects Loss-of-Coolant Accident Drywell pressurization Vent system thrust and pressurization loading Downward air bubble pressure load NMP Unit 1 UFSAR Section VI VI-5 Rev. 25, October 2017 Pool swell liquid impact and drag loads Upward air compression load Submerged structure loads - velocity and acceleration drag loads Froth impingement loads Pool fallback loads Postswell wave loads Steam flow condensation loads on torus walls Lateral condensation loads on downcomers Containment design pressure loads Drywell depressurization Asymmetrical effects Pool thermal stratification effects 3.0 Plant-Unique Modifications As a result of the Mark I Containment Program, many modifications have been performed at Nine Mile Point Nuclear Station - Unit 1 (Unit 1). These include: Y-Quenchers Vent header deflectors Downcomer tie straps Saddles Catwalk removal Relief valve vacuum breakers Torus attached piping Resupport of relief valve discharge lines NMP Unit 1 UFSAR Section VI VI-6 Rev. 25, October 2017 B. PRIMARY CONTAINMENT - PRESSURE SUPPRESSION SYSTEM 1.0 Design Bases 1.1 General The pressure suppression containment system houses the reactor vessel, reactor coolant recirculation loops, and their branch connections. The system is generally considered to extend to the first isolation valve on lines which penetrate the containment. Certain lines which connect to, or penetrate, the containment are designed as containment extensions. These lines function after containment isolation, such as the core spray, containment spray and emergency cooling systems. These systems are discussed in detail in Section V and VII. 1.2 Design Basis Accident (DBA) The specific accident employed as a containment design basis is an instantaneous rupture of the reactor coolant system (RCS) corresponding to a double-ended break of the largest pipe in the containment (coolant recirculation line). An analysis of ruptures over the full range of pipe sizes is included in Section XV and Section VII-A, where it is demonstrated that the rupture of the largest pipe imposes the most severe duty on the containment. The design temperature and pressure of the containment structures are based on the peak conditions following blowdown from the largest pipe rupture. As water and steam flash into the drywell, most of the nitrogen gas in that vessel is rapidly (approximately 10 sec) displaced to the suppression chamber through ten large-diameter vent pipes. Since the suppression chamber vent to atmosphere is closed, a 22-psig backpressure occurs with the rapid addition of the drywell nitrogen gas to that initially present in the gas space of the suppression chamber. The model used for the mixture flashing from the rupture(1) compares favorably with test results at Moss Landing. The calculations for transient pressures and temperatures employed a compressible flow model also based on the Moss Landing tests, as reported on pages III-10 to III-13 of the First Supplement to the PHSR (Preliminary Hazards Summary Report). Initially, the containment is assumed to be at normal temperature and pressure operating limits. The drywell is at 150°F, the suppression chamber at 90°F. Both chambers have an internal pressure of 2 psig. The only heat transfer mechanism considered during blowdown is condensation of steam in the suppression chamber NMP Unit 1 UFSAR Section VI VI-7 Rev. 25, October 2017 water. Condensation on the containment walls and components is conservatively neglected after blowdown. The containment spray system is initiated, condensing steam and cooling the suppression chamber water. The peak pressure in the drywell would be about 34 psig, unless the large rupture was preceded by a small leak which would have prepurged the drywell of nitrogen. In this case the peak drywell pressure could reach about 50 psig. The higher peak drywell pressure results from simultaneous combination of the maximum backpressure in the suppression chamber, displacement of the water leg in the vent pipes and the maximum flow friction losses. Following termination of blowdown the two chambers rapidly equilibrate at about 22 psig. This latter pressure is due mainly to nitrogen gas pressure in the suppression chamber with only a small contribution from vapor pressure. The containment is designed for simultaneous occurrence of the design internal pressure and temperature, plus the dead, live, and seismic loads imposed on the shell. A detailed analysis of the loads imposed on the containment and the resulting stresses is included in Section XVI. With the core spray system in operation, the metal-water reaction is negligible and a peak suppression chamber pressure of 22 psig occurs immediately after blowdown before the containment spray system is initiated. In the highly unlikely event that the core spray system does not operate, a peak suppression chamber pressure of 25 psig occurs following a substantial metal-water reaction. The occurrence of a metal-water reaction does not affect the peak drywell pressure. The 62 psig design pressure for the drywell includes a substantial margin over the pressure requirements discussed above and confirmed by the Moss Landing tests. These tests indicate that for a break of the largest pipe with previous purging of the drywell nitrogen to the suppression chamber, the maximum pressure would be 50 psig. Thus, the drywell design includes a safety margin of 12 psi. The suppression chamber at 35 psig design pressure also includes a substantial safety margin of 10 psi over the calculated peak pressure. During the blowdown period (10 to 15 sec), about 310 million Btu of primary coolant system internal energy are released as steam and water. Subsequently, core decay heat and any energy contributions from the metal-water reaction are released to the containment. The amount of chemical energy released depends on the extent of metal-water reaction. For the maximum reaction, 73 million Btu of energy are released. NMP Unit 1 UFSAR Section VI VI-8 Rev. 25, October 2017 Following the blowdown, the core spray system removes decay heat and any chemical energy from the core. This system draws water from the suppression chamber and sprays over the top of the core (see Section VII-A). Some of this water is converted to steam which is released to the drywell. The maximum steam release rate to the drywell after blowdown is about 50 lb/sec, corresponding to all the decay heat and chemical energy from the maximum metal-water reaction. One containment spray loop consisting of one spray pump, heat exchanger, and raw water pump is capable of condensing this steam and removing all the above heat from the containment. (Currently, to satisfy Appendix J, Type C requirements, a water seal is maintained at the containment spray piping and suppression chamber penetrations at 110 percent of maximum containment pressure, which requires two-pump operation. See Sections VI-D.1.1 and VII-B.) The core spray system will preclude any significant metal-water reaction. However, the containment spray system is designed (with redundancy) to accommodate a substantial metal-water reaction and subsequent hydrogen release associated with complete meltdown of the core. These capabilities are discussed in Section VII. With both core and containment sprays in operation, the pressure and temperature conditions after blowdown are less severe than during blowdown, even if a substantial metal-water reaction occurs. The containment is filled with nitrogen during operation to preclude a flammable hydrogen-oxygen atmosphere following a metal-water reaction. This system is also described in Section VII. The parameters used in the containment design are consistent with experimental data developed for the Bodega Bay pressure suppression containment system by Pacific Gas and Electric Company at its Moss Landing Test Facility. A parameter comparison with Moss Landing tests was given in the First Supplement to the PHSR, page III-11. As stated on page III-6 of the PHSR, provisions are made to accommodate the jet forces resulting from pipe breaks inside the drywell. The effects of safety valve actuation have been investigated, and precautions have been taken to guard against missile hazards and other dynamic effects which result from pipe breaks. An extensive analysis of jet and missile forces is included in Section XVI. 1.3 Containment Heat Removal NMP Unit 1 UFSAR Section VI VI-9 Rev. 25, October 2017 Normal operating temperature and pressure in the pressure suppression system are maintained within specified limits to prevent possible overpressurization during the DBA. Pressure is maintained by the containment inerting system described in Section VII-G.2.0. Drywell coolers are used to maintain temperature, as described in Section VI-E. Provisions are made to purge contaminated nitrogen through the emergency ventilation system to the stack. For long-term, post-accident cooling, the containment spray system is used. This system, with only one spray pump and one raw water pump operating, is capable of removing all the decay heat and chemical energy as described above. (Currently, to satisfy Appendix J, Type C requirements, a water seal is maintained at the containment spray piping and suppression chamber penetrations at 110 percent of maximum containment pressure, which requires two-pump operation. See Sections VI-D.1.1 and VII-B.) 1.4 Isolation Criteria A system of isolation valves completes the containment by blocking inlet and outlet lines of the drywell and suppression chamber when containment isolation is required. These valves are described in Section VI-D. 1.5 Vacuum Relief Criteria As steam is condensed in the drywell by the containment spray, the pressure drops below suppression chamber pressure. If the pressure in the suppression chamber is more than 8.94 psi above drywell pressure, internal components such as the vent pipes and headers could collapse. In order to keep this differential below the maximum allowable of 8.94 psi, relief valves are provided between the two chambers. These valves equalize drywell and suppression chamber pressure, thus preventing a backflow of water from the suppression pool into the vent header system. As steam is condensed and the containment is cooled, pressure drops. If the pressure drops below atmospheric, the drywell and suppression chamber could rupture. To prevent containment pressure from dropping below the vacuum rating of 2 psi for the drywell and 1 psi for the suppression chamber, relief valves are provided between the suppression chamber and the atmosphere. 1.6 Flooding Criteria NMP Unit 1 UFSAR Section VI VI-10 Rev. 25, October 2017 The containment spray raw water intertie, discussed in Sections VII-A and B, can be used to flood the containment above core level using lake water (assuming a source of ac power is available). For long-term post-accident recovery, provision is also made to flood the containment to above core level using the injection sources described in the severe accident procedures (SAP). 1.7 Shielding Inside the drywell, a reactor shield wall surrounds the reactor vessel. This serves during operation to prevent overheating of the biological shield surrounding the drywell. During shutdown, the reactor shield wall attenuates gamma rays, permitting work inside the drywell. (See also Section XII-B.) 2.0 Structure Design 2.1 General The drywell, as shown on Figure VI-1, is a steel pressure vessel with a 70-ft diameter spherical lower portion and a 33-ft diameter cylindrical upper portion. The cylindrical section has a bolted cover. The pressure suppression chamber is a steel pressure vessel in the shape of a torus, below and encircling the drywell. The principal design parameters for the system are summarized below. Suppression Drywell and Vents Chamber Total Volume (No Equipment) 242,700 cu ft 209,000 cu ft Approximate Free Volume 180,000 cu ft 120,000 cu ft Internal Design Pressure 62 psig 35 psig Internal Design Temperature (Maximum) 310°F 205°F Design Leakage NMP Unit 1 UFSAR Section VI VI-11 Rev. 25, October 2017 Rate at Design Pressure 0.5 w/o* per day 0.5 w/o* per day External Design Pressure 2 psig 1 psig The containment system is entirely within the reactor building and, consequently, is not subject to wind or snow loads. The containment was designed according to seismic criteria as a Class I structure, as described in Section III and, as such, the reaction of the design earthquake was added to all dead and live loads carried by the drywell shell. The drywell and suppression chamber were designed, erected, and tested by the Chicago Bridge & Iron Company under a contract with Niagara Mohawk Power Corporation (NMPC). The vessels were designed and constructed in accordance with Section III, Class B, of the ASME Code, and each vessel was stamped with the ASME symbol for the internal pressure and temperature. Code approval calculations are included in Section XVI. Special precautions not required by Codes were taken in the fabrication of the steel drywell and suppression chamber shells. For example, the plate was preheated to a minimum temperature of 200°F prior to welding of all seams thicker than 1 1/4 in, regardless of surrounding air temperature. Preheat at a minimum of 100°F was applied prior to welding of all seams 1 1/4 in or less in thickness when the ambient temperature fell below 40°F. Charpy V-notch specimens were used for impact testing of plate and forging material to give assurance of proper material properties. A detailed analysis of the stresses resulting from various combinations of loads, test results, and a review of inspection procedures are included in Section XVI. The suppression chamber supports transmit vertical and seismic loading to the reinforced concrete foundation slab of the reactor building. Space is provided outside the chamber for inspection and maintenance. Ten large vent pipes form a connection between the drywell and the suppression chamber. The pipes are enclosed with sleeves and are provided with expansion joints to accommodate differential motion between the drywell and suppression chamber. Projecting downward from the header are 120 downcomer pipes, terminating about 4 ft below the water Weight percent NMP Unit 1 UFSAR Section VI VI-12 Rev. 25, October 2017 surface of the pool. The vent header has the same temperature and pressure design requirements as the vent pipes. 2.2 Penetrations and Access Openings High-integrity piping penetrations, access openings, and air locks are provided throughout to maintain the integrity of the containment system. There are four basic types of penetrations: electrical, hot fluid lines, cold fluid lines, and lines welded directly to the containment. The penetrations have the same or higher design pressure and temperature as the containment. Details of typical electrical penetrations are shown on Figures VI-2 and VI-3. These penetrations can be pressurized to design pressure by an inner and outer seal arrangement, allowing verification of tightness. For fluid lines with temperatures above 150°F, penetration details are shown on Figure VI-4. Figure VI-4 shows the typical penetration arrangement. Figure VI-4A shows the configuration for the main steam penetrations (X-2A and X-2B) which have been repaired by adding "clam-shell" bellows joints over the existing joints, which have been removed. These penetrations have a guard pipe between the hot line and the penetration attachment to the drywell steel in addition to the double-seal arrangement. In this manner the penetration is protected against overpressurization should the hot line rupture inside the penetration. The hot fluid from a rupture of this type would be vented into the drywell by the guard pipe. The guard pipes are designed to the same pressure and temperature as the fluid line. The hot fluid penetrations have two expansion bellows, one inside the drywell and one outside. These are designed to take the thermal expansion as well as any movement due to a line rupture. As a final precaution, the inner bellows are designed for a lower pressure, 62 psig, than the outside bellows, 120 psig, to assure inward leakage in the event of failure. The pipe sleeve which attaches to the drywell is designed for 62 psig but, because of structural thicknesses, the sleeve can withstand substantially higher pressures. The only lines which connect to a high-pressure system and do not have a double-seal penetration sleeve are the hydraulic lines to the control rod drives (CRD). These involve 262 small stainless steel lines shopwelded to the three sections of the drywell plate. The mechanical problems involved with this number of small penetrations in a relatively small area make it impractical to provide individual penetration sleeves. The NMP Unit 1 UFSAR Section VI VI-13 Rev. 25, October 2017 pipes are designed to deflect with the drywell shell. They are not individually testable, but will be tested as part of the overall containment leak rate test. For fluid lines with temperatures below 150°F, penetration details are shown on Figure VI-5. These also have double-sealed arrangements, but have no guard piping between the line and the penetration attachment to the drywell. The pipe sleeve which attaches to the drywell is designed for 62 psig but because of structural thicknesses the sleeve can withstand substantially higher pressure. No bellows are required since thermal expansion is minimal. As with all double-ended penetrations, the inner penetration seal is designed to fail at a lower pressure than the outer seal on the pipe sleeve. Lines which open directly to the containment do not have a separate penetration sleeve and are welded directly to the containment shell, e.g., containment pressure measurement, and containment spray connections to the suppression chamber. Lines that open directly to the containment may also be installed in and welded to a penetration sleeve; if a line is installed in a penetration sleeve, the sleeve is considered to be an extension of containment. In addition to the drywell head, two double-door air locks and one bolted hatch are provided for access to the drywell. The locking mechanisms on each air lock door are designed to maintain a tight seal when the locks are subjected to internal pressure. The doors are mechanically interlocked; neither door may be operated unless the other door is closed and locked. When one door is opened, a locking mechanism is latched so that the shafts and gears controlling the other door cannot be moved. The drywell head is bolted and double sealed by two O-ring gaskets. The bolted hatch is also double sealed by two O-ring gaskets. The space between the O-rings can be pressurized to check for leakage. The suppression chamber has three bolted access hatches. These have double O-ring gaskets on the sealing surface. The space between O-rings can be pressurized to check for leakage. 2.3 Jet and Missile Protection An air space averaging about 2 1/2 in is provided outside the drywell between the steel shell plate and the biological shield concrete. This air space is included to allow for drywell deflection due to earthquake-induced relative motions between the drywell and surrounding concrete at a time when the thermal and pressure deflection on the vessel are at a maximum due to the blowdown. NMP Unit 1 UFSAR Section VI VI-14 Rev. 25, October 2017 The air space also will limit drywell shell deformation in the unlikely event it should be struck by either a small missile or a fluid jet arising as a result of certain types of accidents inside the drywell. Studies were conducted to scope the magnitude of credible missile and jet forces which could be encountered. The results of these studies indicate that drywell integrity would be maintained in all cases with the limited amount of deflection which can occur. These studies are summarized in Section XVI. A dished and stiffened jet deflector plate is provided at the drywell opening to each of the vent pipes leading to the suppression chamber to prevent possible damage to the vents from missiles or jet forces. The free-flow area around the periphery of the deflector is about 1.5 times the vent cross-sectional area to minimize flow losses. 2.4 Materials All pressure plates of the drywell are constructed of A201 and A212 Grade B (Firebox) steel made to A300 requirements. The thickness of the steel ranges from about 0.7 to 1.1 in in most of the drywell shell. The neck between the spherical and cylindrical parts of the drywell is 2 5/8 in. This neck section and the drywell head flange were stress relieved because of the thickness. The suppression chamber (torus) is constructed of A201 Grade B (Firebox) steel plates with a certified minimum thickness of 0.46 in. A corrosion allowance of 1/16 in was originally added to the calculated thicknesses of all pressure parts of the drywell, vents and suppression chamber. However, the addition of LOCA and SRV hydrodynamic loads to the containment design bases subsequently reduced this original corrosion allowance. Specifically, engineering analysis demonstrates that for the added LOCA and SRV hydrodynamic loads, the minimum wall thickness required in the most highly stressed portion of the torus shell is 0.431 in. The NRC has reviewed this analysis and found it acceptable (see Section VI-G.2). A program has been developed to monitor the torus shell material thickness for thinning due to corrosion. The program consists of the following: NMP Unit 1 UFSAR Section VI VI-15 Rev. 25, October 2017 1. Ultrasonic (UT) thickness measurements of torus bays are performed periodically. 2. Corrosion sample coupons with the same steel material as that of the torus shell are installed at the water line in the suppression pool (approximately one-half above and one-half below the water line). The corrosion rate obtained from the corrosion coupons is compared to that from the UT measurements, and the most conservative corrosion rate is used to make future corrosion rate determinations. This monitoring program assures that the torus shell material will not be reduced to less than the minimum required wall thickness in any future operation. 2.5 Shielding A reactor shield wall surrounds the reactor, attenuating thermal neutrons and gamma flux to the containment shell. This shield is a hollow cylinder of concrete 2 ft thick, supported on the same structural concrete as the reactor vessel. The portion of the shield in the vicinity of the reactor core is high-density concrete while the remaining portions are ordinary concrete. The inside and outside surfaces of the concrete are covered with steel plate and reinforcing steel is used in the concrete to give added structural strength. The shield is cooled on both surfaces with cool air recirculating from the drywell cooling system. 2.6 Vacuum Relief The following valves are supplied for relief between the drywell and suppression chamber and between the suppression chamber and atmosphere. 1. Between the Drywell and the Suppression Chamber Air Space Number of valves 4 Relief capacity per valve, lb/sec 240 P at indicated capacity, psi 3.3 2. Suppression Chamber Relief to Atmosphere NMP Unit 1 UFSAR Section VI VI-16 Rev. 25, October 2017 Number of valve sets* 3 Relief capacity per set, lb/sec 70 P at indicated capacity, psi 1.0 The P values are for design purposes to provide a reference level for the valve capacity. The actual differentials during the worst-case accident with all valves opening are, at most, 1 to 2 psi (from the pressure suppression chamber to the drywell), and 0.7 psi (from the atmosphere to the suppression chamber). The valves between the vessels are swing-check valves. Each set of vacuum relief valves to the atmosphere consists of a check valve and an air cylinder-operated butterfly valve piped in series with the check valve. The solenoid-actuated air-powered butterfly valve is tripped by either of two vacuum switches which sense the pressure in the pressure suppression chamber relative to atmosphere. 2.7 Containment Flooding Two systems are provided to manually introduce water into the containment at the Operator's discretion after an accident. The containment spray raw water intertie to the containment spray and core spray systems can supply lake water to flood the containment (discussed in Sections VII-A and VII-B). Additional injection systems are available to flood the containment as specified in the SAPs. The drywell is vented through the upper vent and purge line. The reactor vent line has motor-operated valves (MOV) which can be manually opened to vent the reactor vessel directly to the drywell. This provision allows the reactor core to be flooded as the containment water level rises. The stresses on the containment structure, resulting from flooding up to about 7 ft below the operating floor, were analyzed and are discussed in Section XVI. C. SECONDARY CONTAINMENT - REACTOR BUILDING 1.0 Design Bases The reactor building completely encloses the pressure suppression system. This structure provides secondary containment when the pressure suppression system is in service, and primary containment, during refueling, maintenance, or One set is two valves in series. NMP Unit 1 UFSAR Section VI VI-17 Rev. 25, October 2017 testing, when the pressure suppression system is open or not required. The major safety function of the secondary containment is to minimize ground-level release of airborne radioactive materials by providing controlled, elevated release of the building atmosphere through a filter system under accident conditions. When the pressure suppression system is in service, the DBA for the reactor building is the same as for the pressure suppression system--the LOCA without core spray (Section XV). When the pressure suppression system is open, the DBA is the most severe refueling accident, as discussed in Section XV. For either accident, an emergency ventilation system with particulate and charcoal filters is used to reduce radioactivity release to the environment. The reactor building is designed for a maximum in-leakage rate of 100 percent of the building volume per day at 0.25 in of water internal vacuum and neutral wind conditions. Under other than neutral wind conditions, reactor building exfiltration could occur as discussed in Section XV. 1.1 Wind and Snow Loadings Exterior loadings for wind, snow and ice, used in the design of the reactor building, meet all applicable codes as a minimum. The roof and its supporting structure are designed to withstand a loading of 40 psf of snow or ice. The walls and building structure are designed to withstand an external loading of 40 psf of surface area, which is approximately equivalent to a wind velocity of 125 mph, 30 ft above ground level. 1.2 Pressure Relief Design Pressure relief is provided to prevent collapse of the superstructure due to a break of an emergency cooling system or other primary coolant system line in the reactor building. Breaks in all primary coolant system piping have been analyzed since accidents of this type result in the highest pressure, temperature and humidity conditions in the building. A break in the emergency cooling system is the most serious since it releases the most coolant at the highest rate. After accounting for steam condensation and heat losses through the building wall, building temperatures can still be as high as 307°F locally for short time periods, and reach approximately 150°F for the entire building for longer periods of time. Based on analysis of primary coolant system line breaks in the reactor NMP Unit 1 UFSAR Section VI VI-18 Rev. 25, October 2017 building, a metal wall area of approximately 2,400 sq ft has been attached with bolts that are designed to fail with an internal pressure of approximately 65 psf of wall area. Relief of pressure through this area in case of an energy release will prevent excessive internal pressure on the superstructure walls, roof and their supports, which would fail at an internal pressure in excess of 80 psf. Subsequent calculations were performed in accordance with the AISC Manual of Steel Construction, Load & Resistance Factor Design (LRFD), First Edition, to compute the failure load of the building superstructure, and was determined to be at least 117 psf. 1.3 Seismic Design The reactor building and its contents whose functional failure could cause significant release of radioactivity, or which are vital to safe shutdown and isolation of the reactor, are designed as Class I structures using the maximum credible earthquake ground motion of 11 percent of gravity. As discussed in Section III, dynamic analyses determine the earthquake acceleration applicable to the various elevations of the reactor building. All equipment whose functional failure could cause significant release of radioactivity, or which are vital to safe shutdown and isolation of the reactor in the reactor building, is designed to withstand these forces. Functional load stresses (normal operation), when combined with stresses due to earthquake loading, are within the established code stresses. 1.4 Shielding The reactor building shielding is discussed in Section XII-B, and is designed to limit the radiation level in accessible areas during power operation. 2.0 Structure Design The reactor building houses the refueling and reactor servicing equipment; fresh and spent fuel storage facilities; and other reactor auxiliary or service equipment, including the emergency cooling system, reactor cleanup system, liquid poison system, CRD hydraulic system equipment, core and containment spray systems, and components of electrical equipment. The equipment arrangement and principal dimensions are shown on Figures III-2 to III-8. NMP Unit 1 UFSAR Section VI VI-19 Rev. 25, October 2017 2.1 General Structural Features The poured-in-place reinforced concrete building substructure is founded on firm Oswego sandstone. The substructure begins 68 ft

• Also see Section XVI, Subsection G. below grade and extends upward 147 ft to the operating floor. The maximum bearing pressure on the rock is 40 tons/sq ft. This results in a safety factor of 18 based on actual unconfined compressive strength tests on selected specimens of rock core extracted from test borings. The maximum actual bearing pressure on the rock is 14 tons/sq ft. The superstructure above the operating floor is 57 ft high and consists of structural steel framing supporting the roof system, insulated metal panel siding, and a 125-ton overhead crane. The reactor building is enclosed from el 193 (68 ft below grade) to el 340 (operating floor level), with poured reinforced concrete walls varying in thickness from 1 ft 4 3/4 in to 4 ft 0 in. The superstructure (approximately 57 ft high) above the operating floor is enclosed with insulated metal wall panels and a metal roof deck covered with a 5-ply tar and felt built-up roof. The metal wall panels of the superstructure use caulking of demonstrated leak-tightness. Air infiltration tests on joints between panels sealed with the caulking were conducted at the Housing Research Laboratory, University of Miami, in accordance with the recommended specifications of the National Association of Architectural Metal Manufacturers, and showed no measurable air leakage at pressure differentials ranging from 0.3 to 2.0 in of water. The metal panels, insulated precast concrete wall panels, and the related caulking materials, doors and access openings have been carefully analyzed to assure that resultant leakage will be within specifications. Precast concrete slabs (el 261 to 285) and uninsulated metal wall panels (el 285 to 340) are applied to the exterior of the reinforced concrete walls of the reactor building for esthetic purposes. However, these slabs and panels do not form a part of the building support or provide any additional measure of leak-tightness for the building. Fiberglass thermal insulation is provided for the superstructure walls. Thermal insulation properties of the concrete walls, metal wall panels and roofing provide very adequate weather and thermal protection for the reactor building. All materials meet NMP Unit 1 UFSAR Section VI VI-20 Rev. 25, October 2017 ASTM specifications and are in accordance with Fire Underwriters' requirements. The exterior of the reactor building below grade is provided with a peripheral drain for collection of groundwater seepage; the drain discharges into a sump pit with two 150-gpm submersible pumps located at the southwest corner of the building. The reactor building grade floor at el 261 is 12 ft above maximum lake level (el 249). Protection of the building from possible inundations, ice accumulation and lake wave action is provided by a rock dike 1,000-ft long at the shoreline. Specific codes that are complied with include American Institute of Steel Construction, American Concrete Institute, New York State Building Code and the Uniform Building Code (UBC). A computer analysis was made to determine the maximum induced seismic accelerations, displacements, shears, moments and reactions acting on the RPV and its support and on the reactor building. The analysis includes the response of the RPV and its support to the design earthquake and jet reaction forces. Also included is the effect on the RPV of the displacement of the reactor building and containment vessel due to the postulated earthquake. Results of this analysis are contained on Figures VI-6 through VI-17. Personnel access into the reactor building is controlled from the track bay extension and from the turbine building. The track bay extension has a railroad entrance and a personnel access air lock passageway from the outside. The track bay extension consists of a 20-ft by 20-ft by 80-ft long air lock, connected to the track bay compartment by a vertical lift inner door and an airtight seal. The track bay extension is equipped with a motor-operated double swing outer door 16 ft wide by 17 ft 6 in high. The door can also be operated manually and is designed to resist an internal or external load of 40 psf. The outer door closes against a closed cell sponge neoprene closure to provide an airtight seal. The inner vertical lift door bears against a one-piece inflatable seal of reinforced ethylene propylene diene monomer around its perimeter. The entire contact area of the inflatable seal will expand approximately 3/4 in under pressure. The seal material will remain pliable and seal at temperatures of -20°F to 210°F. Containment integrity for the track bay compartment and extension is provided by an outside double swing door, an inside NMP Unit 1 UFSAR Section VI VI-21 Rev. 25, October 2017 vertical lift door and personnel doors connected by an airtight access passageway. The track bay compartment (with extension) and its access openings are shown on Figure III-4. Typical door seals for the personnel and equipment doors are shown on Figure VI-18. Interior doors with air locks are provided in the south wall of the reactor building leading into the turbine room at el 261, as shown on Figure III-4, and at el 340, as shown on Figure III-8. The doors of the air lock have neoprene seals with sealing requirements equivalent to those of the railroad door. Details are shown on Figure VI-19. Procedures and alarms are used to control access and maintain building integrity. Primary and secondary shielding is discussed in Section XII. D. CONTAINMENT ISOLATION SYSTEM 1.0 Design Bases Isolation valves are provided on lines penetrating the drywell and pressure suppression chamber to assure integrity of the containment when required during emergency and post-accident periods. Isolation valves which must be closed to assure containment integrity immediately after a major accident are automatically controlled by the reactor protection system (RPS) described in Section VIII. The drywell and suppression chamber penetrations are dedicated to specific purposes as shown in Tables VI-1 and VI-2, respectively. The tables list the number, size, and type of penetration associated with each purpose. Containment isolation valves (also called isolation valves) are defined as any valves which are relied upon to perform a containment isolation function on lines penetrating the primary reactor containment and include all reactor coolant isolation valves and all primary containment isolation valves. Test, vent and drain (TVD) valves located on the containment pressure boundary are containment isolation valves but are not included in the tables of reactor coolant isolation valves or primary containment isolation valves. Reactor coolant isolation valves are containment isolation valves which are on lines penetrating the primary reactor containment and are connected to the RCS (or a system containing NMP Unit 1 UFSAR Section VI VI-22 Rev. 25, October 2017 reactor coolant) and function as reactor coolant pressure boundary (RCPB) components. Reactor coolant isolation valves function as primary containment isolation valves in the event of a LOCA. Primary containment isolation valves are containment isolation valves on lines penetrating the primary reactor containment connecting directly to the free space enclosed by the containment. Table VI-3a is a listing of all reactor coolant isolation valves, and Table VI-3b lists primary containment isolation valves. All lines which are part of the RCPB and penetrate the primary reactor containment are provided with redundant isolation valves. As a general rule, one of each pair of isolation valves in series is located inside the containment. The other valve is outside the containment. On the emergency cooling system supply and on the feedwater system where it was necessary to install both valves outside the containment, a guard pipe is installed between the line and the containment vessel penetration sleeve. This sleeve is welded to the body of the first isolation valve outside the containment. This, in effect, extends the containment to include the body of the first isolation valve. For the emergency cooling system supply, the two valve bodies are welded end to end for greater integrity. For the feedwater system, the two valves are separated by a 10-in extension. Lines which are part of the reactor coolant boundary and may be required to have flow after an accident are provided with check valves. The CRD and liquid poison systems have two check valves in series. One valve is inside the containment. The feedwater system, as described above, has two valves outside the containment, one of which is a check valve. The cleanup and shutdown cooling systems each have redundant isolation valves with one valve inside the containment. The outer valve on the return to the reactor line is a check valve. Post-accident thermal overpressurization protection is provided for the penetration piping between the isolation valves in the shutdown cooling system. For each emergency cooling system condensate return line penetration, the primary containment isolation function is accomplished either by (a) Type C leak rate testing both the inboard and outboard isolation valves, in which case the NMP Unit 1 UFSAR Section VI VI-23 Rev. 25, October 2017 emergency cooling closed loop outside containment (CLOC)configuration does not apply, or (b) Type C leak-rate-testing one valve and the closed system piping outside primary containment. In this case CLOC configuration applies. The closed system boundary includes the emergency cooling piping and connected branch lines up to and including the first branch line isolation valve. Although in the case of emergency cooling system CLOC configuration, only one valve and a closed system outside containment are credited for the emergency cooling system condensate return lines, these lines still include a second isolation valve. To reduce radiation exposure to personnel, in a CLOC configuration only one of the two valves in each penetration is local leak rate tested per 10CFR50 Appendix J and credited for leakage control. Instrument lines are provided with redundant valving outside the containment. Automatic flow check valves minimize loss of reactor coolant in the event of an instrument line break. All external isolation valves are located as close to the containment as possible. Where guard pipes are used between the containment penetration and the line, the outer valve is welded to the guard pipe. For reactor coolant isolation valves on low-temperature lines where no guard pipe is required, the outer valve is welded directly to the penetrations sleeve. Most lines which connect directly to the containment atmosphere and penetrate the primary reactor containment are provided with redundant isolation valves. Two normally-closed valves outside the containment are provided for systems which are not required to function under accident conditions. Lines which are not equipped with double isolation valves have been determined to be acceptable based upon the fact that the system reliability is not compromised, the system is closed outside containment, and a single active failure can be accommodated with only one isolation valve in the line. Instrument lines connected to containment atmosphere which penetrate primary containment are provided with two isolation barriers, such as manual valves, caps, or diaphragm assemblies. Each containment spray line which is required to be open under accident conditions contains a check valve outside the containment. These check valves are installed to minimize bypassing of pressure suppression during the initial pressure transient of the LOCA.

NMP Unit 1 UFSAR Section VI VI-24 Rev. 25, October 2017 The oxygen sample return line and the nitrogen purge line for the traveling in-core probes use two check valves in series outside the containment. The traveling in-core probe guide tubes use a ball valve and manually-actuated explosive shear valve in series outside containment. Each line that penetrates primary reactor containment and is neither part of the RCPB nor connected directly to the containment atmosphere, in the case of the drywell cooling and recirculation pump cooling systems, has one isolation valve. These systems circulate cooling water in a closed system into and out of the containment. Each line carrying incoming cooling water is provided with a self-actuating check valve outside the containment. Each line which carries water out of the containment has a MOV which is actuated by remote manual control. The isolation system for each line is designed to accommodate loss of power to an isolation valve. MOVs (ac or dc) are designed to fail in the mode in which they are when loss of power occurs. Air-operated valves (AOV) fail closed upon loss of power. Different power sources for each valve in series ensure that the isolation function will not be defeated by single failure. Failure of a single power source does not prevent isolation even where a normally open MOV fails open. Isolation is effected either by having a closed piping system which does not communicate with containment atmosphere or by having a redundant separately powered valve in series with the failed valve. In the case of systems which are required to be open following an accident, valves are normally open and fail open, are normally closed but fail open, or are normally closed but fail closed (as is) but have a redundant valve path in parallel that is open and/or fails open. Systems which connect to the nuclear steam supply system (NSSS) and may be required to have flow after an accident are provided either with two check valves or a check and a remote manually controlled valve in series. These are the feedwater, the CRD hydraulic, and the liquid poison systems. Instrument lines that run from the reactor primary system through the drywell are equipped with shutoff valves and a flow check valve located outside containment as indicated on Figure VI-20. The flow check valves meet or exceed the following design requirements: NMP Unit 1 UFSAR Section VI VI-25 Rev. 25, October 2017 Design Conditions Operating Pressure 1250 psig Operating Temperature 575°F Specified Flow to Close Valve 25 gpm Horizontal Acceleration 0.20 g Vertical Acceleration 0.10 g A cross section of a typical 3/4-in check valve is shown on Figure VI-21. The valve poppet is held open by the spring. The force generated by the pressure differential over the seat area acts against the spring. Flow creates a pressure differential which overcomes the spring and closes the poppet. The differential pressure then acts on the poppet seating area to keep the poppet closed. A bypass arrangement is used on these instrument lines as a means of equalizing line pressure to open the flow check valve in the event it should close and for blowdown purposes. Instrument line leaks can be detected by one or a combination of the following: 1. Operator comparing readings with several instruments monitoring the same process variable such as reactor level, recirculation pump flow, steam flow, and steam pressure. 2. By annunciation of the control function, either high or low in the control room. 3. By a general increase in the area radiation monitor readings throughout the reactor building. 4. By audible noise either inside the turbine building or outside the reactor building. 5. By alarms on the reactor building floor drain tank. 6. By probable increase in area temperature monitor readings in the reactor building. NMP Unit 1 UFSAR Section VI VI-26 Rev. 25, October 2017 Routine surveillance as indicated in items 1 through 6 is felt to be a sufficient program for the periodic testing and examination of the valves in these small-diameter instrument lines. At each major refueling outage, each instrument line flow check valve will be tested for operability. The engineered safeguards systems which may be required to operate following an accident originally had no specific isolation requirements. These systems, which consist of core and containment spray and the emergency cooling system, were designed as containment extensions and diligent efforts were made to meet the intent of Section III-1965 of the ASME Code. Valves were provided in the lines from the suppression chamber and in those into the drywell to provide system isolation for maintenance or testing. Isolation valves for these systems are shown in Tables VI-3a and VI-3b. The opening times, failure modes, and normal position of the valves in the core spray, containment spray and emergency cooling systems are based on the individual system operational requirements as discussed in Sections V and VII. In general, the closure time of all isolation valves is such that the release of fission products to the environment is minimized. As described in Section XV, no large-scale fission product release occurs before 1 min has elapsed. The valve closure times are thus set for a 1-min maximum unless operational restrictions are more severe. The closure times of all valves on lines in systems connecting to the NSSS are based on preventing fuel damage from overheating with no feedwater makeup following a line break in the particular system. The valve closure time for the main steam line (MSL) is based on the MSL break accident discussed in Section XV. By keeping the valve closure time less than about 10 sec, sufficient coolant will remain in the reactor vessel to provide adequate core cooling. The valves are designed to close and to be leak-tight during the worst conditions of pressure, temperature and steam flow following a break in the MSL outside the pressure suppression system. The codes used in the design of Class I system containment isolation valves at the time of construction included ASME Section I-1962 or ANSI B31.1-1955 and ANSI B16.5-1955, with requirements of ASME Section III-1965 for nondestructive testing (NDT). For subsequent modifications, Regulatory Guide (RG) 1.29 recommendations are followed. Piping system segments penetrating containment and considered susceptible to thermal NMP Unit 1 UFSAR Section VI VI-27 Rev. 25, October 2017 overpressurization are analyzed in accordance with the criteria of the ASME Boiler & Pressure Vessel Code, Section III, Appendix F (1986 Edition). The design criteria for containment isolation valves consist of normal and special loadings, load combinations, and load combination limits. Seismic design criteria are listed in Table VI-4. 1.1 Containment Spray Appendix J Water Seal Requirements Table VI-3b lists primary containment isolation valves of the containment spray system which enter the free space of the containment. These lines have an Appendix J water seal by virtue of system operation following the design basis LOCA. The system design basis is continuous operation following the DBA as documented in Section XV-5.3. The Boiling Water Reactor Owners' Group (BWROG) Emergency Procedure and Severe Accident Guidelines (EPG/SAG) restrict drywell and suppression chamber spray operation. The emergency operating procedures (EOP), based on the BWROG EPG/SAG spray limitations, are intended to provide Operator guidance to prevent beyond design basis evaporative cooling conditions from developing. The evaluation, which determined the impact of the EOP assumed actions upon the licensing basis, concluded that the radiological impact of the potential leakage from the primary containment for the conditions where the water seal is secured would result in less than 20 percent of the 10CFR100 regulatory limits, and less than 65 percent of the control room regulatory limits per 10CFR50 Appendix A, General Design Criterion (GDC) 19. The drywell spray limitations were developed to address evaporative cooling conditions which are beyond the Unit 1 design basis. Therefore, the conditions which interrupt the 10CFR50 Appendix J water seal are evaluated as beyond design basis conditions. In this respect, the maximum potential leakage assumed in this evaluation is not included as part of the design basis primary containment leakage. The leakage is only used to compare the maximum potential leakage relative to 10CFR100 and 10CFR50 Appendix A, GDC 19. In order to ensure that assumptions used in this evaluation remain valid, surveillance tests are required to monitor packing degradation and ensure minimal system cross-tie leakage (see Section VI-F.1.2 and VII-B.4.0). Post-accident secondary containment conditions are defined based on the integrity of the containment spray system pressure NMP Unit 1 UFSAR Section VI VI-28 Rev. 25, October 2017 boundary and the containment isolation check valves. The secondary containment conditions are defined based on total leakage of 1.5 percent per day as defined in Section VI-F. This is based on the integrated leak rate testing (ILRT) discussed in Section VI-F.1.2. Therefore, post-accident equipment qualification conditions or post-accident vital area access is not affected by the potential leakage used to evaluate the beyond design basis EOP conditions which terminate the containment sprays. 2.0 System Design A list of all isolation valves on lines penetrating the containment vessels and their pertinent modes and characteristics is given in Table VI-3. Figure VI-22 shows all valves, except those on instrumentation lines. On in-flowing lines, either of two valve arrangements is used. Either both isolation valves in series are self-actuated check valves, one inside and one outside the containment, or one is a check valve and the other is a power-operated valve (electric motor or air). On lines where flow may be in either direction, both valves are power operated. Motive power for each of a pair of power-operated isolation valves in series is from physically independent sources to preclude the possibility of a single malfunction interrupting power to both valves. AOVs which close for the normal containment isolation mode fail closed on loss of motive power. Electric MOVs fail as-is. The power-operated isolation valves are closed automatically from selected signals of the RPS or other sources, as described in Section VIII. Other control valves (such as in the core spray system) are normally closed and open automatically upon appropriate signal after an accident. As a backup to the automatic signals, certain power-operated valves may be operated manually from the main control room. All containment isolation valves, including their power operators, were designed to operate under the most extreme ambient conditions of pressure, temperature, etc., to which they may be exposed after a major accident. An environmental qualification (EQ) program for electrical equipment has been conducted in accordance with 10CFR50.49. As a result of this program, all electrical equipment in the containment isolation system important to safety has been qualified to operate in the NMP Unit 1 UFSAR Section VI VI-29 Rev. 25, October 2017 environment to which it is exposed. Isolation valves in lines connecting to the NSSS and all pipe-welded connections were fully radiographed to assure their integrity. They were built to the applicable ASME Codes and all nuclear interpretations applying to these codes. All power operators on valves inside the containment are ac, since these are considered more reliable than the dc power-operated valves. The piping between containment isolation valves in the drywell floor drain system, the core spray system, the shutdown cooling system, and the post-accident sampling system (PASS) recirculation sample line (penetrations X-25, X-13A, X-14, X-7, X-8, and X-139) is equipped with thermal overpressure protection due to post-accident heatup of the isolated penetrations. The reactor building serves as secondary containment when the reactor is hot (>215°F) and pressurized, and the pressure suppression system is in service. The reactor building ventilation system is provided with two isolation valves in series in both the supply and exhaust ducts. These valves automatically close from a signal of high radiation in the reactor building. If the isolation valves are in an abnormal position during any mode of operation, this abnormality is displayed in the control room on an isolation valve mimic. The containment isolation system design provides safety grade diversity in the parameters needed for the initiation of containment isolation. Containment isolation (except for primary coolant isolation) is initiated on either: 1. High drywell pressure, or 2. Low-low reactor vessel water level. Primary coolant isolation of the main steam, cleanup and shutdown cooling is not initiated on high drywell pressure. Low water level, which may result in fuel failures and, thus, abnormally high levels of radioactivity, will initiate primary coolant isolation. The drywell and suppression chamber H2O2 sampling and containment airborne activity monitor systems isolate on a containment isolation signal. These systems are provided with overrides so that they can be manually reopened for controlled monitoring purposes. NMP Unit 1 UFSAR Section VI VI-30 Rev. 25, October 2017 With a containment isolation signal present, traversing in-core probe (TIP) systems that are in service will switch to reverse manual mode. The TIP probes will automatically retract into their chamber shield and the isolation valves will close. The valves will be secured by the Operator to prevent reopening after the isolation signal clears. In addition to a containment isolation signal, the drywell and suppression chamber vent and purge valves isolate on high radiation at the main stack monitor. The recirculation sample and suppression chamber to waste system lines each have valves capable of automatic isolation on either low-low water level or high drywell pressure. The design of the control system for automatic containment isolation valves is such that resetting or clearing a containment isolation signal will not result in the automatic reopening of containment isolation valves. Opening of containment isolation valves requires deliberate Operator action. 3.0 Tests and Inspections Surveillance requirements for containment isolation valves are given in sections 4.2.7 and 4.3.4 of the Technical Specifications. E. CONTAINMENT VENTILATION SYSTEM 1.0 Primary Containment 1.1 Design Bases During normal Station operation, heat is released to the drywell as heat losses from the reactor, motors, hot pipes and other equipment. The drywell is equipped with six water-cooled heat exchanger fan units, which remove heat generated within the drywell and maintain ambient temperature below 150°F during normal operation, to protect equipment not required under accident conditions. During normal operation, the primary containment vessels--drywell and pressure suppression chamber--are purged with nitrogen (to a pressure of approximately 1.5 psig--see Section VI-B) to maintain less than a 4-percent oxygen concentration, NMP Unit 1 UFSAR Section VI VI-31 Rev. 25, October 2017 which renders the containment atmosphere nonflammable in the event of hydrogen release from a metal-water reaction following a LOCA. 1.2 System Design A piping and instrumentation diagram (P&ID) for the ventilation system is shown on Figure VI-23. The six heat exchanger fan units used to cool the drywell are rated at 500,000 Btu/hr each. Their combined rated cooling capacity is sufficient to maintain the drywell ambient temperature at 135°F maximum with all units in operation, and 150°F maximum with one unit out of service (OOS). The 150°F is the design basis maximum temperature limit for the drywell bulk ambient temperature under normal operation. The minimum drywell bulk average temperature limit is 115°F. This value is adopted as the minimum to bound the acceptable temperature range in which all analysis results and equipment performance is assured. The minimum drywell bulk average temperature is based on providing acceptable accuracy of RPV level instrumentation and to ensure compliance with reactor head safety valve setpoint tolerance, as well as its use as an input to EOP calculations. The coolers are located in the lower portion of the drywell, with ductwork provided on the suction side of the units to draw warmer gases from the upper section of the drywell. The cooled gas is discharged into the lower part of the drywell. The reactor building closed loop cooling water (RBCLCW) system supplies water to the heat exchangers. The RBCLCW system has a backup cooling water supply from the emergency service water (ESW) system. Three coolers can be powered from each diesel generator. The drywell ambient temperature is indicated in the main control room. Since the cooling system forms a closed loop inside the containment, only one isolation valve is included, as discussed in Section VI-D. The design of the combustible gas control system (CGCS) is discussed in Section VII-G. 2.0 Secondary Containment 2.1 Design Bases NMP Unit 1 UFSAR Section VI VI-32 Rev. 25, October 2017 The reactor building has two separate ventilation systems. One system is used during normal operation while the other is a standby to be used under accident conditions. The normal ventilation system supply fans ordinarily provide filtered air to various parts of the building at a rate of approximately one change per hour. However, when conditions warrant, a higher speed operating mode is available which will provide approximately two changes per hour. Unit coolers using service water are installed locally where necessary for additional cooling. During normal operation the pressure inside the building is held slightly negative, approximately 0.25 in W.G. relative to the outside to minimize out-leakage. Exhaust fans discharge all ventilating air to the stack. Exhaust ductwork is arranged to draw air from areas where contamination is most likely to occur, thus preventing its spread into relatively cleaner areas. Both supply and exhaust ducts are provided with two quick-closing leak-tight valves in series which trip closed automatically on high radiation level signal within the building. The emergency ventilation system, to be used under conditions when the reactor building is isolated, is discussed in Section VII-H. 2.2 System Design A P&ID for normal and emergency ventilation is shown on Figure VI-24. The normal ventilation flow rates are about 35,000 and 70,000 cfm, respectively, for normal flow and high-speed purging. The appropriate rate is manually selected by means of a two-speed control on the ventilation fan motors. The air supply equipment consists of a fresh air intake, filter, electric heating units which will automatically control to a set temperature, and two full-capacity fans equipped with inlet vane dampers. Since either fan is capable of the 70,000 cfm rate, one will normally be a full-capacity standby. Supply ductwork with dampers is provided to distribute the air to various areas throughout the building. Two 70,000-cfm exhaust fans (one normally on standby) are provided with connecting ductwork which draws the air mainly from areas of highest potential contamination and exhausts to the stack. NMP Unit 1 UFSAR Section VI VI-33 Rev. 25, October 2017 Local electric heating units and service water-cooled heat exchanger units are provided throughout the building as required to maintain desired temperatures for personnel comfort and equipment protection. The local electric heating units and service water-cooled heat exchanger units are designed to maintain temperatures at 85°F maximum and 70°F minimum in accessible areas and 100°F maximum and 50°F minimum in inaccessible areas. Both the main supply and exhaust ducts are equipped with two leak-tight isolation valves in series, which close automatically upon detection of high radiation levels within the building. The supply and exhaust fans trip immediately. The closure sequence of the normal supply and exhaust isolation valves ensures that reactor building negative pressure is maintained during the transition from normal to emergency ventilation for events which are not accompanied by a loss of offsite power (LOOP) (see Section XV for LOCA/LOOP discussion). They also may be controlled manually from the main control room. The inlet and outlet duct penetrations through the building walls are sealed against leakage. A steel pipe sleeve is integrally cast in the concrete, and the outer end of the sleeve has a gasketed flange which connects to the first isolation valve. F. TEST AND INSPECTIONS A program of testing the primary containment system has been developed based on Appendix J of 10CFR50, "Reactor Containment Leakage Testing for Water Cooled Power Reactors." The program includes overall ILRTs, local leakage rate tests and isolation valve leakage tests. 1.0 Drywell and Suppression Chamber 1.1 Preoperational Testing Following construction of the drywell and suppression chamber, each was pressure tested at 1.15 times its design pressure. Penetrations were sealed with welded end caps as were the downcomers from the drywell to the suppression chamber. The relief lines from the suppression chamber to the drywell were also blanked. Following the strength test, the drywell and suppression chamber were tested for leakage rate at design pressure; each met the criterion for leakage at this stage of construction of less than 0.1 percent per day at design pressure. The suppression chamber was also tested while half filled with water to simulate operating conditions. NMP Unit 1 UFSAR Section VI VI-34 Rev. 25, October 2017 After complete installation of all penetrations, ILRTs of the drywell, suppression chamber, and associated penetrations were conducted. The tests were conducted at several test pressures up to and including 35 psig to establish a leakage rate curve. The necessary instrumentation was installed in the containment systems to provide the data to calculate the leakage rate. Table VI-5 summarizes the initial preoperational tests conducted. 1.2 Postoperational Testing An integrated leakage rate Type A test is performed to demonstrate that leakage through the primary containment and systems and components penetrating primary containment does not exceed the allowable leakage rate specified in the Technical Specifications. The integrated leakage Type A test shall include the containment spray system piping in its operating mode Technical Specification configuration. This is based on the analysis discussed in Section VI-D.1.1 which takes credit for the ILRT as confirmation of containment spray system integrity (i.e., minor components are leak-tight, cross-tie leakage is minimal). The integrated leakage test is conducted at the analyzed maximum accident pressure, Pa. The test pressure, as required by 10CFR50 Appendix J, is based on the design basis LOCA conditions. The peak primary containment pressure following a LOCA would be 35 psig. The Appendix J to 10CFR50 acceptance criteria states that the maximum allowable leakage rate shall not exceed 1.5 weight percent of the contained air in 24 hr at 35 psig. The allowable operational leakage rate shall not exceed 75 percent of 1.0 La and shall be met prior to resumption of power operation following a test. Periodic ILRTs are performed during refueling and maintenance outages in accordance with the performance-based schedule defined in the NMP1 Appendix J Testing Program Plan. The test duration is at least 8 hr. Local testing is performed in accordance with the performance-based schedule defined in the NMP1 Appendix J Testing Program Plan. In addition, whenever double-gasketed seals are opened and closed they are retested. NMP Unit 1 UFSAR Section VI VI-35 Rev. 25, October 2017 2.0 Containment Penetrations and Isolation Valves 2.1 Penetration and Valve Leakage Primary containment testable penetrations and isolation valves required to be Type B or Type C tested are tested at a pressure of at least 35 psig in accordance with the performance-based schedule defined in the NMP1 Appendix J Testing Program Plan. Personnel air lock doors are tested at a pressure of at least 35 psig in accordance with the schedule defined in the NMP1 Appendix J Testing Program Plan. The inner door of the air lock is designed for 62 psig containment internal pressure. However, it is not capable of withstanding the force exerted on it by pressurizing the air lock between the two doors. During preoperational and during periodic testing, the doors are subjected to a test pressure of at least 35 psig with the use of a strongback. Those penetrations and valves which require repairs following an integrated test to meet the allowable leakage rate are subject to local leak tests at a pressure of 35 psig. The main steam isolation valves (MSIV) are tested at a pressure of at least 35 psig at intervals not to exceed 30 months. The combined leakage of all testable penetrations and valves is limited to 60 percent of the maximum allowable leakage rate. The maximum allowable leakage rate (La) is 1.5 percent per day at a pressure of 35 psig (Pa). 2.2 Valve Operability Test All power-operated isolation valves were tested prior to plant initial operation. These valves are presently tested in accordance with the In-Service Testing (IST) Program and Technical Specifications. This testing includes stroke time testing. 3.0 Containment Ventilation System The six drywell coolers (water-cooled heat exchanger fan units) were checked initially and thereafter at each major refueling outage for leakage under normal cooling water pressure. During normal operation, the temperature indicators in the drywell monitor the effectiveness of the coolers. Instrumentation is provided which continuously indicates oxygen content. If the oxygen content increases above 4 percent, NMP Unit 1 UFSAR Section VI VI-36 Rev. 25, October 2017 additional nitrogen is added to reduce oxygen concentration. The relief valves and the overpressure regulators in the nitrogen makeup supply line which prevent the containment from being overpressurized were tested prior to initial startup for operability and setpoint. The purge fans and associated shutoff valves which discharge gas from the containment to the stack were checked for proper operation initially. 4.0 Other Containment Tests System operability tests of the containment spray system are discussed in Section VII. These tests are designed to simulate post-accident conditions and to verify automatic system initiation. 5.0 Reactor Building The reactor building leakage rate is tested by building isolation and operation of the emergency ventilation system. The emergency ventilation system flow control valve is adjusted to obtain 0.25 in of water negative building pressure relative to atmosphere. The rate at which air is exhausted through the system, measured by the flow indicator, indicates building in-leakage. Reactor building in-leakage is related to exfiltration and wind speed, as discussed in Section XV. The reactor building leakage rate is tested each cycle. 5.1 Reactor Building Normal Ventilation System The ventilation system was initially tested to determine that proper distribution of air throughout the building had been obtained. The supply and exhaust line dampers are adjusted for proper pressure distribution to ensure that the flow is from areas of low contamination to areas of high contamination. Flow switches in the supply and exhaust lines provide for low flow alarms in the control room. 5.2 Reactor Building Isolation Valves The isolation valves associated with the reactor building ventilation system are tested for leakage rate as part of the reactor building leakage test discussed earlier. NMP Unit 1 UFSAR Section VI VI-37 Rev. 25, October 2017 Periodic surveillance of the supply and exhaust valves closure times is performed to ensure that the valves close in the proper sequence. 5.3 Emergency Ventilation System Tests and inspections of the emergency ventilation system are discussed in detail in Section VII. G. REFERENCES 1. F. J. Moody, "Maximum Flow Rate of a Single Component, Two Phase Mixture," Journal of Heat Transfer, Trans. ASME, Series "C" Vol. 07, February 1965, p 134. 2. NRC Safety Evaluation, "Approval of Reduction Factors for Condensate Oscillation Loads in Nine Mile Point Nuclear Station Unit No. 1 (NMP1) Torus," August 11, 1994 (TAC No. M85003). NMP Unit 1 UFSAR Section VI VI-38 Rev. 25, October 2017 TABLE VI-1 DRYWELL PENETRATIONS Purpose No. Size Type Control Rod Drive Hydraulic 262 1" Pipe Sleeve Piping Spare 23 3" Pipe Sleeve Spare 33 4" Pipe Sleeve Oxygen Analyzer 3 4" Pipe Sleeve Recirculation Pump Differential 5 4" Pipe Sleeve Pressure Impulse Line Recirculation Flow Transmitter 5 4" Pipe Sleeve Impulse Line Recirculation Pump Seal Pressure 5 4" Pipe Sleeve Impulse Line Emergency Cooling Elbow Flow 4 4" Pipe Sleeve Meter Drywell Pressure 2 4" Pipe Sleeve Reactor Level 11 4" Pipe Sleeve Leak Rate Monitoring 1 4" Pipe Sleeve Main Steam Flow Impulse Line 2 4" Pipe Sleeve Breathing Air 1 4" Pipe Sleeve Service Water 1 4" Pipe Sleeve Containment Sampling 1 4" Pipe Sleeve Core Differential Pressure 1 4" Pipe Sleeve Impulse Line Spare 9 6" Pipe Sleeve In-core Calibration Tubes 4 6" Pipe Sleeve (In Use) In-core Calibration Tubes 3 6" Pipe Sleeve (Spare) Drywell Floor Drain Sump 1 6" Pipe Sleeve Drywell Equipment Drain Sump 1 6" Pipe Sleeve Core Spray Vent 1 6" Pipe Sleeve Spare 5 8" Pipe Sleeve Liquid Poison 1 8" Pipe Sleeve Control Rod Drive System Exhaust 1 8" Pipe Sleeve to Reactor Containment Sampling 2 8" Pipe Sleeve Reactor Level-Triple Low 1 8" Pipe Sleeve Containment Range Radiation 1 8" Pipe Sleeve Monitor Spare 5 10" Pipe Sleeve Recirculation Pump Cooling Water 1 10" Pipe Sleeve Supply Drywell Level Impulse Line 1 10" Pipe Sleeve Recirculation Pump Cooling Water 1 10" Pipe Sleeve Return NMP Unit 1 UFSAR Section VI VI-39 Rev. 25, October 2017 TABLE VI-1 (Cont'd.) Purpose No. Size Type Containment Range Radiation 1 10" Pipe Sleeve Monitor Electrical (In Use) 50 12" Pipe Sleeve Electrical (Spare) 9 12" Pipe Sleeve Containment Spray 4 18" Pipe Sleeve Cleanup Supply 1 22" Pipe Sleeve Closed Loop Cooling Supply to 1 22" Pipe Sleeve Drywell Air Coolers Core Spray 2 22" Pipe Sleeve Closed Loop Cooling Return From 1 22" Pipe Sleeve Drywell Air Coolers Cleanup Return 1 22" Pipe Sleeve Spare 3 22" Pipe Sleeve Emergency Cooling Steam Supply 2 24" Pipe Sleeve Emergency Cooling Return 2 24" Pipe Sleeve Drywell Access Hole 1 24" Pipe Sleeve Drywell Air Vent and Fill 1 24" Pipe Sleeve Shutdown Cooling Supply 1 26" Pipe Sleeve Shutdown Cooling Return 1 26" Pipe Sleeve Vacuum Breaker 4 30" Pipe Sleeve Feedwater 2 30" Pipe Sleeve Main Steam 2 38" Pipe Sleeve Emergency Escape Lock 1 66" Pipe Sleeve Vent Pipe to Suppression Chamber 10 81-1/4" Pipe Sleeve Equipment Lock 1 120" Pipe Sleeve Equipment & Personnel Lock 1 122" Pipe Sleeve Drywell N2 Vent & Fill 1 20" Pipe Sleeve Containment Vent - Post-LOCA 1 3" Pipe Sleeve Sample Line 1 3" Pipe Sleeve Sample Line 1 4" Pipe Sleeve NMP Unit 1 UFSAR Section VI VI-40 Rev. 25, October 2017 TABLE VI-2 SUPPRESSION CHAMBER PENETRATIONS Purpose No. Size Type Suppression Chamber Pressure 1 1" Pipe Sleeve Indicator Spare 3 1" Pipe Sleeve Suppression Chamber Water Level 4 1-1/2" Pipe Sleeve Transmitter Suppression Chamber Cooling 2 4" Pipe Sleeve Spare 2 4" Pipe Sleeve Containment Spray - Pump Test 1 8" Pipe Sleeve Core Spray - Pump Test 2 8" Pipe Sleeve Hydrogen-Oxygen [Monitor (CAD)] 1 8" Pipe Sleeve Suppression Chamber Makeup Water 1 8" Pipe Sleeve Oxygen Analyzer Sample and 1 8" Pipe Sleeve Return Suppression Chamber Makeup 1 8" Pipe Sleeve (Spare) Spare 2 8" Pipe Sleeve Containment Spray Suction 4 12" Pipe Sleeve Core Spray Suction 4 12" Pipe Sleeve Electrical 2 12" Pipe Sleeve Suppression Water Temperature 2 12" Pipe Sleeve Indicator Spare 4 12" Pipe Sleeve Main Steam Relief Valve 6 14" Pipe Sleeve Discharge Suppression Chamber Vent and 1 20" Pipe Sleeve Purge Supply Suppression Chamber Vent and 1 20" Pipe Sleeve Purge Exhaust Cleanup System Relief Valve 1 20" Pipe Sleeve Discharge Vacuum Breaker 4 30" Pipe Sleeve Access Manhole 3 36" Pipe Sleeve Vent Pipe From Drywell 10 81-1/4" Pipe Sleeve Torus Drains 3 4" Pipe Sleeve Torus Water Quality Return Line 1 4" Pipe Sleeve NMP Unit 1 UFSAR Section VI VI-41 Rev. 25, October 2017 TABLE VI-3a REACTOR COOLANT SYSTEM ISOLATION VALVES Line or System No. of Valves (Each Line) Location(9) Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Main Steam(1) (Two Lines) 1 1 Inside Outside 01-01,02 01-03,04 Open Open As Is Closed AC Motor Pn/DC Solenoid 10 10 Close Close Reactor water level low-low, or low reactor pressure (with mode switch in run), or main steam line high flow, or low-low-low condenser vacuum, or high temperature in the steam tunnel Feedwater(1) (Two Lines) 1 1 Outside Outside 31-07,08 31-01R,02R Open Open As Is -- AC Motor Self Act. Ck. 60 -- -- -- Remote manual -- Emergency Cooling Steam Leaving Reactor(1) (Two Lines) 1 1 Outside Outside 39-09R,10R 39-07R,08R Open Open As Is As Is AC Motor DC Motor 38 38 Close Close High emergency cooling system flow Condensate Return to Reactor(1)(10) (Two Lines) 1 1 Inside Outside 39-03,04 39-05,06 Closed Closed -- Open Self Act. Ck. Pn/DC Solenoid -- 60 60 -- Close Open -- High emergency cooling system flow Reactor water level low-low or high reactor pressure NMP Unit 1 UFSAR Section VI VI-42 Rev. 25, October 2017 TABLE VI-3a (Cont'd.) Line or System No. of Valves (Each Line) Location(9) Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Reactor Cleanup Water Leaving Reactor(1) (One Line) Water Return to Reactor(1) (One Line) 1 1 1 1 Inside Outside Inside Outside 33-02R 33-04 33-01R 33-03 Open Open Open Open As Is As Is As Is -- AC Motor DC Motor AC Motor Self Act. Ck. 18 18 18 -- Close Close Close -- Reactor water level low-low, or high area temperature or liquid poison initiation -- Shutdown Cooling Water Leaving Reactor(4)(7) (One Line) Water Leaving Reactor(8) (CIV Bypass Line) Water Return to Reactor(4)(7) (One Line) Water Seal Parallel Valves (See Note 7) 1 1 1 1 1 4 Inside Outside Inside Inside Outside Outside Outside Outside Outside 38-01 38-02 38-216 38-13 38-12 38-169 38-170 38-171 38-172 Closed Closed Closed Closed Closed Closed Closed Closed Closed As Is As Is -- As Is -- -- -- -- -- AC Motor DC Motor Self Act. Ck. AC Motor Self Act. Ck. Self Act. Ck. Self Act. Ck. Self Act. Ck. Self Act. Ck. 40 40 -- 40 -- -- -- -- -- Close Close -- Close -- -- -- -- -- Reactor water level low-low, or high area temperature -- -- -- -- -- -- Liquid Poison(1) (One Line) 1 1 Inside Outside 42.1-02 42.1-03 Closed Closed -- -- Self Act. Ck. Self Act. Ck. -- -- -- -- -- -- Control Rod Drive Hydraulic(1) (One Line) 1 1 Inside Outside 44.3-13 44.3-12 Open Open -- -- Self Act. Ck. Self Act. Ck. -- -- -- -- -- -- NMP Unit 1 UFSAR Section VI VI-43 Rev. 25, October 2017 TABLE VI-3a (Cont'd.) Line or System No. of Valves (Each Line) Location(9) Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Core Spray Core Spray Injection(3) (Two Lines) Core Spray Penetration(3) (Piping Thermal Relief) Core Spray High Point Vent(4) (Two Lines) Core Spray Condensate Supply (Keep Fill)(5) (Two Lines) 2 1 1 1 1 2 Inside Outside Inside Inside Outside Outside 40-01,09, 10,11 40-02,12 40-80 40-83 40-30,31 40-32,33 40-20,21, 22,23 Closed Open Closed Closed Closed Open As Is As Is -- As Is Closed -- AC Motor AC Motor Self Act. Ck. AC Motor Pn/DC Solenoid Self Act. Ck. 22.5 22.5 -- 27 27 -- Open Open -- Close Close -- Reactor water level low-low or high drywell pressure coincident with reactor vessel pressure less than 365 psig Reactor water level low-low or high drywell pressure -- Core Spray System Valves(5) (Two Lines) Core Spray Pump Discharge(4) (Two Test Lines to Suppression Chamber) 1 1 Outside Outside 40-03,13 40-05,06 Closed Closed -- As Is Self Act. Ck. AC Motor -- 27 -- Close -- Reactor water level low-low or high drywell pressure Scram Discharge Volume System Vent**(1) (One Line) 2 Outside 44.2-15,16 Open Closed Pn/AC Solenoid 10 Close Automatic or manual reactor scram Scram Discharge Volume System Drain**(1) (One Line) 2 Outside 44.2-17,18 Open Closed Pn/AC Solenoid 10 Close Automatic or manual reactor scram NMP Unit 1 UFSAR Section VI VI-44 Rev. 25, October 2017 TABLE VI-3a (Cont'd.) Line or System No. of Valves (Each Line) Location(9) Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Post-accident Reactor Sampling(1)(6) (One Line) Reactor Recirculation System Sampling(1) (One Line) 1 1 1 1 Outside Outside Inside Outside 44.1-07 122-03 110-127 110-128 Open Closed Closed Closed -- Closed As Is As Is Self Act. Flow Fuse Pn/DC Solenoid AC Motor DC Motor -- 30 20 20 -- Close Close Close -- Reactor water level low-low or low-low-low condenser vacuum or reactor low pressure (with mode switch in RUN) or high temperature in the steam tunnel or main steam line high flow NOTES:

• Pn - Pneumatically Operated. ** Technical Specification Section 3.1.1e for LCO requirements. (1) These valves do not have to be vented during the Type A test. However, Type C leakage from these valves is added to the Type A test results, if not vented. (2) Deleted. (3) The inside core spray injection isolation valves are water sealed during and after an accident. These valves are leak rate tested with water in accordance with the Appendix J Program. Under 10CFR50, Appendix J, Option B, through RG 1.163, water-sealed CIV test frequency may be set using a performance basis in a manner similar to that described in NEI 94-01, Revision 0, dated 7/26/95, for Type B and Type C test intervals. The outside core spray injection isolation valves are open with their breakers locked in the OFF position. Therefore, the outside core spray injection valves do not have to be tested under the IST or Appendix J Leakage Program. (4) These valves are provided with a water seal. Valves shall be tested consistent with Appendix J water seal testing requirements. Under 10CFR50, Appendix J, Option B, through RG 1.163, water-sealed CIV test frequency may be set using a performance basis in a manner similar to that described in NEI 94-01, Revision 0, dated 7/26/95, for Type B and Type C test intervals. Leakage rates shall be conservatively limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm. (5) These valves are tested in accordance with Technical Specification Section 4.2.7.1a. (6) The self-actuating flow fuse is tested in accordance with Technical Specification Section 4.3.4c. (7) Two 1" globe valves (38-206 and 208) are provided outside in the seal water (core spray) flow test line and one 3/4" globe valve (38-209) is provided outside in the seal water supply line drain, which also serve as RCS isolation valves. (8) -216) is provided inside primary containment around isolation valve 38-01. This valve is provided with a water seal and tested under the Appendix J program for limited flow in the open direction, and under the IST Program, exercised closed for isolation capability. (9) Reactor coolant isolation valves function as primary containment isolation valves in the event of a LOCA. (10) For each emergency cooling system condensate return line penetration, isolation function is accomplished either by (a) Type C leak rate testing both the inboard and outboard isolation valves, in which case the emergency cooling closed loop outside containment (CLOC)configuration does not apply, or (b) Type C leak-rate-testing one valve and the closed system piping outside primary containment. In this case CLOC configuration applies. This meets 10 CFR 50 Appendix J requirements. The closed system boundary includes the emergency cooling system main process piping and connected branch lines up to and including the first branch line isolation valve.

NMP Unit 1 UFSAR Section VI VI-45 Rev. 25, October 2017 TABLE VI-3b PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Line or System No. of Valves (Each Line) Location Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Drywell Vent & Purge N2 Connection (One Line) Air Connection (One Line) 1 1 1 1 Outside Outside Outside Outside 201-32 201-31 201-10 201-09 Closed Closed Closed Closed Closed As Is Closed As Is Pn/DC Solenoid AC Motor Pn/DC Solenoid AC Motor 15 30 15 30 Close Close Close Close Reactor water level low-low or drywell high pressure or high radiation at stack monitoring Suppression Chamber Vent & Purge N2 Connection (One Line) Air Connection (One Line) 1 1 1 1 Outside Outside Outside Outside 201-16 201-17 201-08 201-07 Closed Closed Closed Closed Closed As Is Closed As Is Pn/DC Solenoid AC Motor Pn/DC Solenoid AC Motor 15 30 15 30 Close Close Close Close Reactor water level low-low or drywell high pressure or high radiation at stack monitoring Drywell N2 Makeup (One Line) 2 Outside 201.2-03,32 Closed Closed Pn/DC Solenoid 60 Close Reactor water level low-low or drywell high pressure Suppression Chamber N2 Makeup (One Line) 2 Outside 201.2-06,33 Closed Closed Pn/DC Solenoid 60 Close Reactor water level low-low or drywell high pressure Drywell Equipment Drain Line(1) (One Line) 1 1 Inside Outside 83.1-09 83.1-10 Open Open As Is Closed AC Motor Pn/DC Solenoid 60 60 Close Close Reactor water level low-low or drywell high pressure Drywell Floor Drain Line(1) (One Line) 1 1 1 Inside Outside Outside 83.1-11 83.1-12 83.1-35 Open Open Closed As Is Closed -- AC Motor Pn/DC Solenoid Self Act. Rel. 60 60 -- Close Close -- Reactor water level low-low or drywell high pressure -- Vacuum Relief Atmosphere to Pressure Suppression System (Three Lines) 1 1 Outside Outside 68-08,09,10 68-05,06,07 Closed Closed Open -- Pn/DC Solenoid Self Act. Ck. 5 -- Open -- Negative pressure relative to atmosphere -- NMP Unit 1 UFSAR Section VI VI-46 Rev. 25, October 2017 TABLE VI-3b (Cont'd.) Line or System No. of Valves (Each Line) Location Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Reactor Cleanup System Relief Valve Discharge(2) (One Line to Suppression Chamber) 2 Outside 63.1-01,02 Closed -- Self Act. Ck. -- -- -- H2-02 #11 Sampling Drywell Supply (One Line) Suppression Chamber Supply (One Line) Drywell Return (One Line) Suppression Chamber Return (One Line) 2 2 2 2 Outside Outside Outside Outside 201.7-01, 02 201.2-110, 111 201.7-10,11 201.2-109, 112 Open Open Open Open Closed Closed Closed Closed Pn/DC Solenoid Pn/DC Solenoid Pn/DC Solenoid Pn/DC Solenoid 60 60 60 60 Close Close Close Close Reactor water level low-low or high drywell pressure H2-02 #12 Sampling Drywell Supply(1) (One Line) Suppression Chamber Supply(1) (One Line) Drywell Return(1) (One Line) Suppression Chamber Return(1) (One Line) 2 2 2 2 Outside Outside Outside Outside 201.2-29, 30 201.2-23,24 201.2-67,68 201.2-70,71 Open Open Open Open Closed Closed -- -- DC Solenoid DC Solenoid Self Act. Ck. Self Act. Ck. 60 60 -- -- Close Close -- -- Reactor water level low-low or high drywell pressure -- -- Core Spray Pump Suction (Four Lines From Suppression Chamber) 1 Outside 81-01,02 81-21,22 Open As Is AC Motor 90 -- Remote manual Pump Discharge(4) (Two Test Lines to Suppression Chamber) 1 Outside 40-05,06 Closed As Is AC Motor 27 Close Reactor water level low-low or high drywell pressure NMP Unit 1 UFSAR Section VI VI-47 Rev. 25, October 2017 TABLE VI-3b (Cont'd.) Line or System No. of Valves (Each Line) Location Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Core Spray (cont'd.) Pump Recirculation (Four Lines Connect to Two Test Lines to Suppression Chamber) 1 Outside 81-241,242, 243,244 Closed -- Self Act. Relief Valve -- -- -- Containment Spray Drywell & Suppression Chamber Common Supply(2) (Four Lines) Drywell Branch(2) (Four Lines) Suppression Chamber Branch(2) (One Branch for Each System) Pump Suction From Suppression Chamber (Four Lines) Containment Spray Test Line to Torus(2) (One Line) Compressed Air Test Line(2) (Two Lines) 1 1 2** 1 1 1 Outside Outside Outside Outside Outside Outside 80-15,16, 35,36 80-17,18, 37,38 80-19,39, 65,66,67,68 80-01,02, 21,22 80-118 80-90,91 Open Closed Closed Open Closed Closed As Is(6) -- -- As Is As Is -- Pn/DC Solenoid Self Act. Ck. Self Act. Ck. AC Motor AC Motor -- 60 -- -- 70 60 -- -- -- -- -- -- -- Remote manual -- -- Remote manual Remote manual Local manual Emergency Cooling Vent to Torus(2) (One Line) 2 Outside 05-05,07 Closed As Is AC Motor -- -- Remote manual NMP Unit 1 UFSAR Section VI VI-48 Rev. 25, October 2017 TABLE VI-3b (Cont'd.) Line or System No. of Valves (Each Line) Location Relative to Primary Containment Valve No. Normal Position Fail Position on Loss of Motive Power or Control Signal Motive Power* Maximum Oper. Time (Sec) Action on Initiating Signal Initiating Signal (All Valves Have Remote Manual Backup) Containment Atmosphere Monitoring Supply Line (One Line) 2 Outside 201.7-08,09 Open Closed Pn/DC Solenoid 60 Close Reactor water level low-low or high drywell pressure Containment Post-LOCA Vent (Two Lines) 2 Outside 201.1-09, 11,14,16 Closed Closed Pn/DC Solenoid 60 Close Reactor water level low-low or high drywell pressure N2 Purge - TIP Indexers(1) (One Line) 2 Outside 201.2-39,40 Open -- Self Act. Ck. -- -- -- Traversing In-core Probe(1) (Four Lines) 1 Outside 36-147,148, 149,150 Closed Closed AC Solenoid 60 Close Reactor water level low-low or high drywell pressure Breathing Air Connection (One Line) 1 1 Inside Outside 114-116 114-114 Locked Closed Closed -- -- -- -- -- -- -- -- Local manual Service Water Connection(1) (One Line) 1 1 Inside Outside 72-480 72-479 Locked Closed Closed -- -- -- -- -- -- -- -- Local manual Recirc. Pump Cooling Water(5) Supply Line Return Line 1 1 Outside Outside 70-93 70-92 Open Open -- As Is Self Act. Ck. DC Motor -- 60 -- -- -- Remote manual Drywell Cooler Water(5) Supply Line Return Line 1 1 Outside Outside 70-95 70-94 Open Open -- As Is Self Act. Ck. DC Motor -- 60 -- -- -- Remote manual Hardened Containment Vent System(7) (One Line) 2 Outside 201.13-71, 74 Closed Closed Pn/DC Solenoid -- -- -- NMP Unit 1 UFSAR Section VI VI-49 Rev. 25, October 2017 TABLE VI-3b (Cont'd.) NOTES:

• Pn - Pneumatically Operated. ** One valve in each separate line and one valve in each common line. (1) These valves do not have to be vented during the Type A test. However, Type C leakage from these valves is added to the Type A test results, if not vented. (2) These valves are provided with a water seal capability. No Appendix J or IST leakage rate testing is required. (3) This note deleted. (4) These valves are provided with a water seal. Valves shall be tested consistent with Appendix J water seal testing requirements. Under 10CFR50, Appendix J, Option B, through RG 1.163, water-sealed CIV test frequency may be set using a performance basis in a manner similar to that described in NEI 94-01, Revision 0, dated 7/29/95, for Type B and Type C test intervals. Leakage rates shall be conservatively limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm. (5) These valves do not meet the requirements of 10CFR50 Appendix J, Section II-H. No testing required. (6) Containment spray isolation valve fails open on loss of electrical (dc) power and fails as-is on loss of air. (7) The HCVS valves are not powered from normal station compressed air or power. The valves will always be closed and isolated from their motive force by locked closed valves. The valves are administratively controlled to remain closed as they are not required to be opened during normal operation or any design basis event. The valves would only be opened by manual operator action at the auxiliary control room and HCVs remote operating station during a beyond design basis event. Therefore, there is no signal to automatically close these valves.

NMP Unit 1 UFSAR Section VI VI-50 Rev. 25, October 2017 TABLE VI-4 SEISMIC DESIGN CRITERIA FOR ISOLATION VALVES Seismic Design System Criteria Main steam, feedwater, core spray, 0.30g H/0.15g V makeup from condensate storage tank, drywell cooling, recirculation pump cooling (supply and return outside drywell), reactor cleanup (return in and outside drywell and supply inside drywell), shutdown cooling (return outside drywell), emergency cooling (inside drywell), liquid poison, containment spray (drywell and torus inlet check valves) Emergency cooling, suppression 0.30g H/0.11g V chamber vent & purge, containment vacuum relief, shutdown cooling (supply and return inside drywell), drywell vent & purge, drywell floor drain, drywell equipment drain Containment spray (drywell inlet) 0.26g H/0.10g V Shutdown cooling (supply outside 0.20g H/0.10g V drywell), reactor cleanup (supply outside drywell), containment spray (pump suction) NMP Unit 1 UFSAR Section VI VI-51 Rev. 25, October 2017 TABLE VI-5 INITIAL TESTS PRIOR TO STATION OPERATION INTEGRATED TEST TYPE OF TEST 1. Integrated Leakage Rate Test a. 24 hr minimum b. Various pressures to 35 psig CONTAINMENT PENETRATIONS TYPE OF TEST 1. Pressure test to design pressure for at least 5 min a. Penetrations b. Access locks c. Flanged openings 2. Leakage rate test at 22 psig for at least 1 hr a. Penetrations b. Access locks ISOLATION VALVES OPEN TO THE FREE SPACE OF THE CONTAINMENT TYPE OF TEST 1. Pressure test to design pressure for at least 5 min 2. Leakage rate test at 22 psig for at least 1 hr 3. Valve operability test a. Automatic closure b. Acceptable closure times ISOLATION VALVES CONNECTED TO THE NUCLEAR STEAM SUPPLY SYSTEM TYPE OF TEST 1. Hydrostatic pressure test at operating pressure 2. Valve operability test a. Automatic closure b. Acceptable closure times NMP Unit 1 UFSAR Section VI VI-52 Rev. 25, October 2017 TABLE VI-5 (Cont'd.) CONTAINMENT SPRAY TYPE OF TEST 1. Water spray test a. Automatic initiation 2. Air test VACUUM RELIEF TYPE OF TEST 1. Automatic initiation of power-operated valve 2. Valve operability of swing check EMERGENCY VENTILATION TYPE OF TEST 1. Isolate the reactor building and automatic start of emergency ventilation 2. Filter removal efficiency a. Particulate b. Charcoal DRYWELL AND SUPPRESSION CHAMBER RGURE Vl-1 UfZIAR Rev. 14 (June 19")

• ELECTRICAL PENETRATIONS-HIGH VOLTAGE i IS ,u...JJL...J.J........u...J'-1..-..-U...LL--..u...J.J.....-U-LU-l, I FIGURE Vl-2 UFSAR Ptev. 14 (June 1916)

( ELECTRICAL PENETRATIONS -LOW VOLTAGE Pll'llWIGH llSID£ FACE Of' Dmll!U IAIHH ILlllD Pll'lllG fUllGE DETALA ROURE Vl-3 UFSAR Rav. 14 (June 1991) PJPE PENETRATIONS -HOT . ... .... .. -*-+ . . . . ..... ' . .... . ,. ':)>., z c ;:::: '"' ... z 8 t; ... .... RGURE Vl-4 UFSAR Rev. 14 (June 1996) CLAMSHELL EXPANSION JOINT _)-------1 .-----L...---------c FIGURE Vl-4a UFSAR Rev. 14 (June 1996) ( \ TEST CONNECTION I 2" PLATE TEST CONNECTION I TYPICAL PENETRATION FOR INSTRUMENT LINES "O ** f. PIPE SLEEVE .. ' . . ,.. . .. TYPICAL PENETRATION FOR COLD PIPES PIPE SLEEVE Q .. *, *._ *. *. 1 I I *-' .. r . ! ... . . . .. . . ... 11 * . .. ... 2" AIR SPACE 3/16 PLATE ' INSIDE FACE / *; OF DRYWELL .' 2" AIR SPACE _,. INSIDE FACE OF DRYWELL ,. / 1/4 PLATE FIGURE Vl-5 Uf&AR Rev. 14 (June 1996) 400 350 ...... 300 I.I-z 0 ...... < > u.J ...J UJ 250 200 0 REACTOR BUILDING DYNAMIC ANALYSIS ACCELERATION EAST-WEST DIRECTION MAS/ K SEC. 2 \ ROOF EL. 395' -0"

• 59.5 -............ \.. WITHOUT ROCKING "' \WITH ROCKING '\ \ " \ ' OPERATING FLOOR EL. 340' -0" ) J G 384 I ' J I I *--546 /, I T J I I o--522 /j I I J, I I ' 525 I I II
• 652 // , ' 'j *---1483 , BASE SLAB EL. 198' -0" MODEL 10 20 30 40 ACCELERATION -PER CENT GRAVITY AGURE V1-I ' " ""-"" .. c-'" ... ....... --' c::> ""-' ...,._ "" ' C'>-.... UFIAR Rev. 14 (.Jun. 1996) 400 350 I-LL.I LL.I 300 z 0 I-< > LL.I _J LL.I 250 200 0 REACTOR BUILDING DYNAMIC ANALYSIS DEFLECTIONS EAST-WEST DIRECTION ROOF EL. 395' -0" I I I I I -V" WITHOUT ROCKING/ -----/ --WITH ROCKING / / II ( OPERATING FLOOR EL. 340' -0" j ' I/ I BASE SLAB EL.198'-0" 100 200 300 400 500 600 700 DEFLECTION -MILS FIGURE Vl-7 UFSAR Rev. 14 {June 1991}

400 \ 350 ..... u.J u.J LL.. z: 0 300 ..... < > u.J -I u.J 250 200 0 *' \ \\ ' -REACTOR BUILDING DYNAMIC ANALYSIS ELEVATION VS. BUILDING SHEAR EAST-WEST DIRECTION ROOF EL. 395' -0!' OPERATING FLOOR EL. 340' -0" "\ "' " " """' \.. " " '" ""' '\ ' "" ., ' I\.. ' " "'""' '\WITH ROCKING .. WITHOUT ROCKING\ ' " \ \ , \ ' BASE SLAB EL. 198' -0" \ \ . 5000 10000 15000 20000 25000 30000 SH EAR* KIPS FIGURE Vl-8 UFSAA Rev. 14 (June 19945) ( \ ...... LL.I LL.I LI.. z: 400 350 0 300 j::: < > LL.I ....J LL.I 250 200 ' 0 \\ '\ \ REACTOR BUILDING DYNAMIC ANALYSIS ELEVATION VS. BUILDING MOMENT EAST-WEST DIRECTION ROOF EL. 395' -0" OPERATING FLOOR EL. 340' -0" \\ \ "' ' "' I" " '" WITHOUT ROCKING"' "" ROCKING ""-' ' ... """ " BASE SLAB EL. 198' -0" -500 1000 1500 2000 2500 3000 MOMENT (FT. -KIPS) THOUSANDS FIGURE Vl-9 UFSAR Rev. 14 (June 1996) ( \ 400 350 .... I.LI I.LI IJ,.. z 300 0 .... < > I.LI _J I.LI 250 200 0 / 10 REACTOR BUILDING DYNAMIC ANALYSIS . ACCELERATION NORTH-SOUTH. DIRECTION ROOF EL. 395' -0" .............. ' ' \ WITHOUT ROCKING\ 1 WITH ROCKING I I I / OPERATING FLOOR EL. 340' -0" I J v I I I , j j I I' I I/ I j I/ II ) ' BASE SLAB EL. 198' -0" 20 30 40 ACCELERATION -PER CENT GRAVITY FIGURE Vl-10 UFSA,ft Rev. 14 (June 1996) ( I-L&J L&J ..... . REACTOR BUILDING DYNAMIC ANALYSIS DEFLECTIONS NORTH-SOUTH DIRECTION 400 ROOF EL 395'

• O" I __,,.--WITHOUT ROCKING // 350 i----t----1/,....,./-+---+--+---+---+--+--+--+--+--+--t-----l-----f---I I OPERATING FLOOR EL. 340' -0" l j . II 300 II ;:::: cc > I L&J ...J L&J 250 ....... 200 a:=::i::======i=+=r:=+::=+=:+:==t.:B:.;,,;A:;;S *,.:.O w--=-;:-=3-1 0 100 200 300 400 500 600 700 DEFLECTION -MILS FIGURE Vl-11 UFIAR Rev. 14 (June 199t) 400 350 300 ..... < > LU -I LU 250 200 0 l\ " \\ \\ \ REACTOR BUILDING DYNAMIC ANALYSIS ELEVATION VS. BUILDING SHEAR NORTH-SOUTH DIRECTION ROOF EL. 395'-0" \ OPERATING FLOOR EL. 340' -0" ' , .. "" "" "" "\ "" "' """ " i\. '\ '\..WITH ROCKING ' \ .. WITHOUT ROCKING'\. \ \. \ \ BASE SLAB EL. 198' -0" '\. 5000 10000 15000 20000 25000 30000 SHEAR* KIPS FIGURE Vl-12 UFSAR Rev. 14 (.June 1996)

( \ 400 350 g 300 < > UJ -..I UJ 250 200 l -0 '\ REACTOR BUILDING DYNAMIC ANALYSIS ELEVATION VS. BUILDING MOMENT NORTH -SOUTH DIRECTION RO OF EL. 395' -0" OPERATING FLOOR EL. 340' -0" \ \\ \ '\ '\ " """ " " ' .. " "'-' """ WITHOUT ROCKING" .... ROCKING I ' I ' BASE SLAB EL. 198' -Ott " ' 500 1000 1500 2000 MOMENT (FT.-KIPS) *THOUSANDS 2500 FIGUR:E Vl-13 3000 UfSAR Rev. 14 (June 1996) 310 300 290 280 I-..... ...... ..... z: 270 0 j:: < > ...... ...J ...... 260 250 240 230 REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. ACCELERATION INCLUDES THE EFFECT OF I BLDG. MOVEMfNT & BASE ROCKING /" -* ' ""' STABILIZER .., REACTOR BLOG * .,; SYSTEM *'. o/ -....I _. < ';.., 311: c .... ...: ....I -LIJ ::c: I .. .... "' -...J .... ...... .... "" , "" I-...... J > = 0 ...J I I-' ...J (,J .... ' < ... 8-< ...... ...: i-!ii: = -Cl /J ...J ...... :c "" f .. *--.... .... I ... ... ...; -I P JINT MAss(K SEC 2) FT. 7 1 33.8 --re 2 39.8 , .I 3 21.0 4 30.4 ... °' .,,; !/ 5 18.2 = -...... 6 20.4 0 z: -, 10.4 ...J >

• 8: (,J -I--ii? = 0 a.. J a.. :::> .. "" <D I °' I BASE EL. 225'
• 6" IMODEI 10 20 30 40 50 60 ACCELERATION -PER CENT GRAVITY FIGURE VI-14 UFSAR Rev. 14 (June 1996) 310 300 290 280 I-UJ UJ u... z: 270 0 j::: < > UJ ....I UJ 260 250 240 230 REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. DEFLECTION I / ' t r....JET THRUST CONO./ v , j f -SEISMIC -' I -' < -31::: Cl ...J ) UJ / fl Q v / Co? r I/ / J I i / INCLUDES THE EFFECT OF BUILDING MOVEMENT J/ r ,-7 -/* .. r.i t.J ,_ Q.. ii? 0 100 200 DEFLECTION (MILLS) FIGURE vt-15 UFIAR Rev. 14 (Ju.rte 19H)

( \ 310 300 290 280 I-w w LL. . z: 0 j::: 270 < > w ...I w 260 250 240 230 REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. SHEAR ------"

• I ' I -I . I I
• i I !
• I { < i\ I /'\. I I ; I ' ;r ----ff} i" r\. I ' I I I 1--* \ I I I \ I I : JET TH RUST , I 'e I I I I I I ' I I I \ I 0 200 400 600 800 1000 SHEAR *KIPS FIGURE Vl-11 UflAR Rev. 14 (June 1191) z: e .... c > w -I w ' REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. MOMENT 310 .. \ STR. EL. 303' .9*1----+---+---+-------' 300
• f"l --1------+--+--+---+---+---+---+---+----l"----+--+--+---1 I \f; ' ,a +-I ""' \ r" I f;\CO. rt--' 280 250 240 230 0 ' I 10 l .,_. SEISMIC-(INCLUDES EFFECT OF BLDG. \ MOVEMENT AND BASE ROCKING) I """ ' ' \ \ \ 1 BASE EL. 225'
• s* 20 30 40 MOMENT (FT.-KIPS) X 103 '-+----1 50 60 FIGURE Vl-17 UFIAlll Rev. 14 (June 1996)

TYPICAL DOOR SEALS 14 6A. STEEL JAMB SEAL FLOOR SEAL A IR. LOCK DOORS INTERIOR SIDE EXTERIOR SIDE CLOSED CELL SPON6E NEOPRENE .. . . * * '4 ... . . " <l "'

• 11-.v .v. " ** v A OUTER BAY DOOR e. C> * * * .... I>
• 4 . . . e. C> * * * .... I>
• 4 ... e. C> * *
• A />
• 4 *** DOOR IN CLOSED POSITION INNER R.R. BAY DOOR FIGURE Vl-11 UFSAR Rev. 14 (June 1196)

.. I .. ' ... .. .. ' .. ' I ..I/£ 1-.1 I I DETAILS OF REACTOR BUILDING AIR LOCKS ..... . . . :,..: L *< .. . .; ... . ... *.-. *; . ... .. ,,. ... ,.._** _____ .::-_ .. . .. :. AGURE Vl-19 UFSAR Plev. 14 (June 19M) INSTRUMENT LINE ISOLATION VALVE ARRANGEMENT TO INSTRUMENTS VALVE 1" GLOBE VALVE VALVE CHECK VALVE VALVE REACTOR VESSEL DRY WELL FIGURE VI-20 UFSAR REVISION 17 OCTOBER 2001 TYPICAL FLOW CHECK VALVE Body Poppet Clip Spring Pin Ft*GURE Vl-21 UFSAR Rev. 14 (June 1996) -* !' a c TJ 0(.1) 0 D .... ::0 0 CJ ::0 co co ' < Ntll '-J J N Ul !SJ 1-RPS COJ11h11IH'4Eltl 5"""' S!ICTIO't TJ CQ'tllll...:HI SPMf CJ Sl.CTIO't c ::0 fTl < 7 N N CQllf9Mr TEST CiEtlERAL NOTES1 ------L.:D:,,RYWELL H0 REACTOR I, THIS IJRAll!Nll 15 SCHEMATIC DH..T TO OEl'ICT lCX:ATIO'i OF COtHAINMENT ISQ..ATION VAt.VES. A\.L ,.Ull<<,,llliL BLOCICltlCl VALVES AND TEST.VENT,& CRAIN Vl'ILVES 11R[ OEl'ICTEO. 2.CONTAINMENT SPRAY 15(l.ATJQN VALVE FAILS IJl'[N ON LOSS OF ELECIRJCAL IDCJ POWER llNO FAILS AS*IS G"I t.055 AIR. * = 1VALVES, [CU!hENI CNlll Ul 0 r D 0 z < D r < fTl Ul -< Ul fT1 3:: SOLENOID VALVES ARE "ELECTRICALLY RETIRED IN DRYWELL COOLING SYSTEM T I I I I . I + I I I I I I I I REACTOR DRYWW. i ,1 ,' ,, , ,: . ,, ,, ' , ,, , .l r -a* x 5' -0" VEHT OPEHING'S TYPICAL.OF 4 r-rxr-r -PASSAGEWAY REACTOR SHIEUl IALL Ir DIA. VEHT OPENING 10 EQUALLY SPACED FIGURE VI-23 UFSAR Rev1s1on 16 November 1999 REACTOR BUILDING VENTILATION SYSTEM llllKYDUCT SERVICE WATER CAVJTY T CONNECTION --+-------< 1 1J.200CFH STDRAGEPIT 1,ilil0CFH ELEVATOR 120ilCFM PIPECHl'lSE 1ee CFM FLDOREL.318' 3500CFH C.U.DEMrn.& POST FILTER 400CFH C.U.FILTER 4a0 CFM INSTRUMEl'IT ""'" see CFM PERSONNEL lEOUIP.LOCK Wllli!I CFM FUELPCJDL FILTER SLUDGE Plf.IP 100 CFM FLOOR EL.237' & 218' 271!111l CFM FLOOR EL.237'& 218' eee CFM I I _ _J FIGURE VI-24 UFSAR Rev1s1on 18 October 2003 U.S. NUCLEAR REGULATORY COMMISSION DOCKET 50-220 LICENSE DPR-63 NINE MILE POINT NUCLEAR STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT (UPDATED) OCTOBER 2017 REVISION 25 NMP UNIT 1 UFSAR Section VII EF VII-1 Rev. 25, October 2017 SECTION VII LIST OF EFFECTIVE FIGURES Figure Number Revision Number VII-1 17 VII-2 17 VII-3 18 VII-4 14 VII-5 14 VII-6 16 VII-7 14 VII-8 14 VII-9 14 VII-10 16 VII-11 14 VII-12 25 VII-13 17 VII-14 19 VII-15 19 VII-16 19 VII-17 22 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section VII VII-i Rev. 25, October 2016 SECTION VII ENGINEERED SAFEGUARDS A. CORE SPRAY SYSTEM 1.0 Design Bases 2.0 System Design 2.1 General 2.2 Operator Assessment 3.0 Design Evaluation 4.0 Tests and Inspections B. CONTAINMENT SPRAY SYSTEM 1.0 Licensing Basis Requirements 1.1 10CFR50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 1.2 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants 2.0 Design Bases 2.1 Design Basis Functional Requirements 2.2 Controlling Parameters 3.0 System Design 3.1 System Function 3.2 System Design Description 3.3 System Design 3.4 Codes and Standards 3.5 System Instrumentation 3.6 System Design Features 4.0 Design Performance Evaluation 4.1 System Performance Analyses 4.2 System Response 4.3 Interdependency With Other Engineered Safeguards Systems 5.0 System Operation 5.1 Limiting Conditions for Operation 6.0 Tests and Inspection C. LIQUID POISON INJECTION SYSTEM 1.0 Design Bases 2.0 System Design 2.1 Operator Assessment 3.0 Design Evaluation 4.0 Tests and Inspections 5.0 Alternate Boron Injection NMP Unit 1 UFSAR Section Title Section VII VII-ii Rev. 25, October 2016 D. CONTROL ROD VELOCITY LIMITER 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 3.1 General 3.2 Design Sensitivity 3.3 Normal Operation 4.0 Tests and Inspections E. CONTROL ROD HOUSING SUPPORT 1.0 Design Bases 2.0 System Design 2.1 Loads and Deflections 3.0 Design Evaluation 4.0 Tests and Inspections F. FLOW RESTRICTORS 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections G. COMBUSTIBLE GAS CONTROL SYSTEM 1.0 Design Bases 2.0 Containment Inerting System 2.1 System Design 2.2 Design Evaluation 3.0 Containment Atmospheric Dilution System 3.1 System Design 3.2 Design Evaluation 4.0 Tests and Inspections H. EMERGENCY VENTILATION SYSTEM 1.0 Design Bases 2.0 System Design 2.1 Operator Assessment 3.0 Design Evaluation 4.0 Tests and Inspections I. HIGH-PRESSURE COOLANT INJECTION 1.0 Design Bases NMP Unit 1 UFSAR Section Title Section VII VII-iii Rev. 25, October 2016 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections J. Gas Accumulation 1.0 Nine Mile Point Response to GL 2008-01 K. REFERENCES NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section VII VII-iii Rev. 25, October 2016 VII-1 PERFORMANCE TESTS NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section VII VII-iv Rev. 25, October 2016 VII-1 CORE SPRAY SYSTEM VII-2 FIGURE DELETED VII-3 CONTAINMENT SPRAY SYSTEM VII-4 thru FIGURES DELETED VII-5 VII-6 LIQUID POISON SYSTEM VII-7 MINIMUM ALLOWABLE SOLUTION TEMPERATURE VII-8 FIGURE DELETED VII-9 TYPICAL CONTROL ROD VELOCITY LIMITER VII-10 CONTROL ROD HOUSING SUPPORT VII-11 HYDROGEN FLAMMABILITY LIMITS VII-12 COMBUSTIBLE GAS CONTROL SYSTEM VII-13 H2-O2 SAMPLING SYSTEM VII-14 CONTROLLED HYDROGEN AND OXYGEN CONCENTRATIONS - INERTED CONTAINMENT VII-15 INTEGRATED POST-LOCA NITROGEN VOLUME REQUIREMENT - INERTED CONTAINMENT VII-16 CONTAINMENT PRESSURE FOLLOWING CAD ACTUATION - INERTED CONTAINMENT VII-17 FEEDWATER DELIVERY CAPABILITY (SHAFT DRIVEN PUMP) TO TIME AFTER TURBINE TRIP FOR 1000 PSIG REACTOR PRESSURE AND 1.0 INCH HG ABS EXHAUST PRESSURE NMP Unit 1 UFSAR Section VII VII-1 Rev. 25, October 2017 SECTION VII ENGINEERED SAFEGUARDS This section deals with several principal design features other than the containment system which have been expressly provided as precautionary safeguards to either prevent or mitigate the consequences of major accidents. The systems to be included in this section are somewhat arbitrarily selected in the sense that many other portions of the Station, including systems of a supporting nature such as emergency power and water supply, might be called upon in the event of an accident. In addition, although the high-pressure coolant injection (HPCI) system is discussed in this section, it is not an engineered safety feature (ESF). The HPCI system is a mode of the feedwater system (FWS). It was designed to provide a reliable source of high-pressure injection in the event of a small line break. However, when the Nine Mile Point Nuclear Station - Unit 1 (Unit 1) emergency core cooling system (ECCS) was analyzed to show compliance with 10 CFR 50.46(b), the HPCI system was not considered because it does not have a source of emergency power. Additional analyses, design and testing reports are presented in Section XVI - Special Topical Reports. A. CORE SPRAY SYSTEM 1.0 Design Bases The core spray system consists of two separate and independent core spray loops to prevent overheating of the fuel following a postulated loss-of-coolant accident (LOCA). Each loop has redundant active components within itself. This system is designed to accommodate the range of LOCAs from the smallest up to the largest line break as discussed in Section XV. An environmental qualification (EQ) program for electrical equipment has been conducted in accordance with 10CFR50.49. As a result of this program, electrical equipment in the core spray important to safety has been qualified to operate in the environment to which it is exposed. 2.0 System Design 2.1 General The core spray system shown on Figure VII-1 is designed in accordance with the ASA B31.1-1955 Piping Code with certain NMP Unit 1 UFSAR Section VII VII-2 Rev. 25, October 2017 requirements of the ASME Code, Section II-B-1965, for the piping and valves. Isolated piping sections considered susceptible to thermal overpressurization are analyzed in accordance with the criteria of the ASME Boiler & Pressure Vessel Code, Section III, Appendix F, 1986 Edition. The pump casings and discharge strainer housings (baskets removed) are designed in accordance with the ASME Code, Section III-B-1965. The system is of welded construction and where flanged connections are necessary double seals are used. Equipment and piping between the suppression chamber and the topping pumps are designed to 310 psig and for a temperature range of -10°F to 205°F. An exception to the 310 psig design pressure for this portion of the system is the core spray motor cooling coils which are designed to 150 psig. Cooling water for the motor coolers flows from discharge side of each core spray pump, through flow limiting orifices, through the motor cooling coils, and then to the suction side of the pump. Relief valves are provided in the cooling water line to provide overpressure protection for the motor cooling coils associated with Core Spray Pump Motors 111, 112 and 122. The relief valve for the motor cooling coil associated with Core Spray Pump Motor 121 has been gagged closed to prevent premature opening, which has been evaluated as acceptable under temporary modification ECP-15-000458. The piping and equipment after the topping pump to the containment system isolation check valve are designed to 470 psig and the same temperature range. The system outside the drywell is also qualified to a maximum torus bulk temperature of 165°F applied to a thermal stress analysis. These pressures include the pump shutoff heads and post-accident design pressure. The equipment and piping (carbon steel) between the suppression chamber and the isolation check valve will not be subjected to reactor pressure and temperature. Safety valves set at 349 psig are utilized for pressure protection of this section of the system. They also provide minimum flow for pump protection before vessel depressurization during small break LOCA. From the isolation check valve to the reactor, the system is fabricated of stainless steel with a design pressure of 1200 psig and design temperature of 575°F. Core spray water comes from the suppression chamber and returns to the suppression chamber after cooling the core; therefore, no additional coolant supplies are necessary. Normal makeup to the suppression chamber pool is from the condensate storage and transfer system. The suppression chamber water is cooled by the containment spray heat exchangers (Section VII-B). The core spray system is capable of maintaining peak cladding temperature (PCT) below 2200°F and local cladding oxidation less than 13 percent (Section XV-C.2.0). NMP Unit 1 UFSAR Section VII VII-3 Rev. 25, October 2017 Each loop has an additional set of full-capacity pumps which are normally idle, but can be operated simultaneously. Each set of pumps, consisting of one core spray pump and one topping pump, delivers suppression chamber water to a separate ring header located inside the reactor vessel directly above the core; the headers spray water directly onto the fuel bundles in a preestablished pattern. Strainers are included on the pump suction side inside the suppression chamber to screen out particulates which could interfere with the discharge of water from the spray nozzles. Adequate net positive suction head (NPSH) to the core spray pumps is provided. Nuclear Regulatory Commission (NRC) Bulletin 96-03 requested implementation of appropriate procedural measures and plant modifications to minimize the potential for clogging of the ECCS suction strainers as a result of debris accumulation from debris generated during a postulated LOCA. Large-capacity ECCS suction strainers have been designed and installed to account for the worst-case generation, transport, and accumulation of post-LOCA debris to assure a sufficient available NPSH to the pumps. In the event of a total loss of the core spray primary water source (a loss of suppression chamber water below the core spray pump suction level), raw lake water can be supplied to the core spray nozzles to provide an alternate source of core cooling through a tie-in with the containment spray raw water system. Pump operation is automatically initiated from low-low reactor water level or high drywell pressure. Internal isolation valves remain closed until low reactor pressure (365 psig) signals them to open. Each set of pumps (one core spray pump and one topping pump) will, as a minimum, deliver core spray sparger flow as shown in Table XV-9a. Following the design basis accident (DBA), reactor pressure rapidly decreases and water level drops below the core. The low-low water level initiates pump operation. However, the valves do not open until the low reactor pressure permissive switches are also tripped. On initiation signal the core spray pump is started, and after a suitable time delay for watering, the topping pump is started. Until the internal isolation valves are opened, a minimum of 480 gpm is recirculated (through a relief valve set at 349 psig) to the suppression chamber to provide the pump cooling. NMP Unit 1 UFSAR Section VII VII-4 Rev. 25, October 2017 For line breaks smaller than 0.30 sq ft, reactor pressure may not decrease rapidly enough to prevent clad overheating if there is no feedwater flow. Therefore, an automatic depressurization system (ADS) is provided to depressurize the reactor so that the low-pressure permissive signal can be sent to the valves and the core spray water admitted to the reactor. In addition, emergency operating procedure (EOP) jumpers are provided to allow throttling of the inboard isolation valves and opening of the test return line isolation valves by defeating valve interlocks and initiation signals. This permits the core spray system to slowly restore level while providing pump protection for extended recirculation. Six solenoid-actuated relief valves, three of which act as backups either individually or together, are provided to depressurize the primary system to approximately 50 psi. The relief valves discharge to the pressure suppression chamber. The signals used for initiation of the ADS are simultaneous low-low-low reactor water level and high drywell pressure sustained for 120 sec. Both signals are used in a one-out-of-two twice logic for initiating action. If one set or any individual valve of a set fails to operate, the logic circuitry is designed to provide for operation of the second set or any individual second valve after a 5-sec time delay. Fire-induced failure in the control complex circuitry will not cause spurious opening of an electromatic valve. A time delay relay in the confirmatory logic prevents actuation of the confirmatory logic as a result of an interruption of reactor protection system (RPS) power. Annunciators for low-low-low water level, high drywell pressure and automatic depressurization trip are located in the control room. The Operator is provided with override push buttons which will permit delay of automatic depressurization. Each time the buttons are pushed, the time delay cycle is repeated. Keylocked ADS inhibit switches are provided which eliminate the need to reset the timer. Both push buttons or both keylocked switches must be operated to defeat ADS initiation. Power for the relief valve solenoid is supplied from the Station batteries. Power for the instrument channels is supplied from RPS busses 11 and 12. Loss of both RPS busses (11 and 12) will not cause actuation of the ADS system. Power for the logic circuit is supplied from 4160-V power boards (PB) 102 and 103 through auxiliary transformers. (This assures power is available to the core spray system during automatic depressurization.) If ac power were lost to either PB 102 or NMP Unit 1 UFSAR Section VII VII-5 Rev. 25, October 2017 103, one logic system would become inoperable, but the other logic system would remain operable and capable of initiating automatic depressurization. Upon receipt of an actuating signal, the four core spray pumps are sequentially started when powered from either the Station reserve power supply or the diesel generators. A line is routed to the suppression chamber from between the isolation check valves and the outer motor-operated valves (MOVs) to permit flow testing. During these tests, the outer MOV is closed and the valve in the test line to the suppression chamber is opened. Pumps can be started and the flow routed to the suppression chamber. The test return line can also be used for extended minimum recirculation flow to support continuous pump set operation, with or without injection, to support throttling of the inboard isolation valves and the shutdown cooling water seal. Each core spray loop has a high-point vent and a keep-full system to allow testing of isolation valves at full power. During normal operation the high-point vent system is isolated. Keep-full system operation will be continuous during normal operation through normal operation of the condensate pumps. Keep-full isolation or loss of keep-full does not result in an inoperable core spray system. Reasonable judgments may be made to assess keep-full functions, as well as reasonable attempts to maintain the piping filled with water. Prior to the quarterly core spray valve operability test and monthly, the two high-point vent valves will be opened. To verify that the core spray system is solid up to the inner isolation valve, a check to see that flow is present to the equipment drain tank will be made. The vent valves will then be closed and the inside core spray valves operated. A seal water supply line originates from the topping pump discharge header in each core spray loop to pressurize and provide a supply of seal water to the shutdown cooling system isolation valves to meet Appendix J. The shutdown cooling system MOVs will be administratively controlled by closing the valves and then removing power during normal reactor operation, except when shutdown cooling system is required to be placed in service. 2.2 Operator Assessment NMP Unit 1 UFSAR Section VII VII-6 Rev. 25, October 2017 All instrumentation and controls necessary for the Operator to assess the operation of the system are located on the main control room panel. Each core spray loop has separate and independent flow indication and header pressure indicators. Isolation valve control switches and position indicator lights are provided for each isolation valve. The isolation valve positions are also indicated on the isolation valve mimic on the main control panel. Each core spray pump and topping pump has its own control switch with indicating lights and motor ammeters on the control room panels. An indication of suppression chamber level and pressure is also provided in addition to alarms for high and low values for each. Each core spray pump has pressure switches to monitor core spray pump suction pressure. The pressure switches will alert the Operator to a clogged suction strainer. Pressure switches on each loop header are outside the drywell; they will alarm on low core spray header pressure if the system has been called upon to operate. These alarms are sounded on the control room annunciator and the Station computer. All sensing instrumentation is in accessible areas and is provided with suitable valving for in-place testing at any time. Differential pressure indicators are installed to monitor the condition of the core spray piping between the reactor vessel wall and the shroud inside the reactor vessel. The instrumentation is designed to provide a control room alarm if the core spray piping between the reactor vessel and shroud suffers a loss of integrity. 3.0 Design Evaluation It is necessary to maintain continuity of core cooling subsequent to a postulated LOCA to prevent fuel damage. This continuity of cooling is emphasized in the design of the core spray system. The design provides, as a minimum, a flow rate as shown in Table XV-9a for each loop. For large breaks, the core spray system can keep the PCT below 2200°F without assistance from the ADS system. From the largest break down to about 0.30 sq ft, the reactor depressurizes NMP Unit 1 UFSAR Section VII VII-7 Rev. 25, October 2017 sufficiently fast for the core spray to achieve rated flow before the cladding begins to melt. Small breaks, i.e., breaks below about 0.30 sq ft, are those which fall outside the range of the core spray system. In the event of such a break, substantial coolant loss could occur from the reactor vessel while it was still at relatively high pressure. The ADS system is provided which, in conjunction with the core spray system, will prevent significant fuel damage for all sized line breaks. The ADS system is capable of depressurizing the vessel either manually or on a simultaneous low-low-low water level and high-pressure signals. The LOCA analysis, described in Section XV-C, shows the capability to maintain adequate core cooling under the entire range of breaks analyzed. Core spray distribution tests were conducted in air with simulated updraft effects and are the basis for flows and nozzle location. A report on these tests is included in the General Electric Company (GE) Report APED-5458. The effect of steam environments on the spray distribution has been evaluated in NEDE-30241. 4.0 Tests and Inspections Each core spray loop was tested initially during preoperational testing with water under full-flow conditions. Data on flows and pressures at various points in the flow lines was obtained. The nozzle spray pattern was observed as far as practical with the reactor head off. Each loop was also operated bypassing the water to the suppression chamber and the corresponding flow and pressure data obtained. Subsequently, the core spray and topping pumps are periodically operated, and the water pumped from the suppression chamber through the appropriate supply lines to the outer system isolation valve, then returned to the suppression chamber. Flow into the reactor vessel is not attempted since this would introduce relatively impure water into the reactor coolant. Data on the flow rate and pressure at various points for each supply loop are obtained for comparison with the previously established normal conditions. Interlocks are provided such that the valve in the test line cannot be opened unless the motor-operated containment system isolation valves both inside and outside the drywell are closed. These valves cannot be reopened until the test valve is closed. The MOVs on the pump discharge lines to the reactor vessel are periodically opened NMP Unit 1 UFSAR Section VII VII-8 Rev. 25, October 2017 fully and the time to open is recorded. These valves shall be fully open within 22.5 sec (valve stroke time) after the signal is given to assure that, under accident conditions, the total delay in achieving full core spray flow is less than 37 sec. The safety valves on the core spray lines outside the second system isolation valve are periodically removed and tested for setpoint, as recommended by the ASME Code, Section III-B-1965. These valves are also containment isolation valves and are subject to Appendix J Type B and C testing. The pumps and valves are tested quarterly by recycling water to the suppression chamber. Testing of the Core Spray pump and valves is conducted per the Surveillance Frequency Control Program implemented by License Amendment 222 for NMP1. One of the tests performed consists of introducing condensate water into the pump suction and testing automatic initiation of the pumps and valves. At least once per month verification is made that the keep-full system piping is filled with water. Once each quarter during the scheduled operability test, the system is visually inspected for leakage, and maintenance is performed as required. For the differential pressure instrumentation that monitors the core spray piping within the reactor vessel, the differential pressure indications are checked once per day, and the instrumentation is tested/calibrated once every three months. B. CONTAINMENT SPRAY SYSTEM 1.0 Licensing Basis Requirements The following regulatory documents are applicable to the containment spray system (CSS) and, in general terms, form the basis on which the system is designed and operated. 1.1 10CFR50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants An EQ program for electrical equipment has been conducted in accordance with 10CFR50.49. Consequently, electrical equipment important to safety in the CSS system has been qualified to operate in the LOCA environment. NMP Unit 1 UFSAR Section VII VII-9 Rev. 25, October 2017 1.2 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants The Technical Supplement to Petition for Conversion from Power Operating License to Full Term Operating License covered the Unit 1 positions relative to the General Design Criteria (GDC). Those portions of the documentation that cover both the description of the requirements and NMPC's positions relative to these requirements, as they pertain directly to the CSS system, have been extracted and are shown below: Criterion 16 Containment Design Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment, and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. A pressure suppression containment system consisting of a drywell, suppression chamber (torus), and interconnecting vent piping is the primary containment for the main coolant system. During normal operation, the reactor building, containing the pressure suppression system, provides a secondary containment barrier. To ensure the integrity of the primary containment, integrated leak tests were performed prior to Station operation and periodically thereafter, as provided in the Technical Specifications. The results demonstrated that the containment met the design leak rate of 0.5 percent per day at a pressure of 35 psig and, therefore, provides an essentially leak-tight barrier. The design basis LOCA was evaluated at the primary containment maximum allowable accident leak rate of 1.5 percent per day at 35 psig. The analysis demonstrates that the offsite doses from this accident would be well within the limits of 10CFR50.67. Criterion 38 Containment Heat Removal A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and maintain them at acceptably low levels. NMP Unit 1 UFSAR Section VII VII-10 Rev. 25, October 2017 Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that, for onsite electric power system operation (assuming offsite power is not available), and for offsite electric power system operation (assuming onsite power is not available), the system safety function can be accomplished. Two CSS system loops are provided to remove heat, reduce pressure, and restore the pressure suppression system temperature following a LOCA. Each loop is capable of removing all the decay heat and, in addition, the energy from any credible metal-water reaction at a rate that will prevent containment pressures and temperatures from exceeding their design values. The power for the pumps is provided from redundant Station reserve power supply systems or from one of two emergency diesel generators. One of the two spray loops is automatically actuated on the combined condition of high drywell pressure and low-low reactor water level. The other loop can be manually controlled from the main control room. Criterion 39 Inspection of Containment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system. Essential CSS system components are inspected periodically to ensure the integrity and capability of the system. The system tests and inspections are described in Section VII-B-6.0 and in the Technical Specifications. Criterion 40 Testing of Containment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic pressure and functional testing to assure: 1) the structural and leak-tight integrity of its components, 2) the operability and performance of the active components of the system, and 3) the operability of the system as a whole, and under conditions as close to the design as practical, the performance of the full operational sequence that brings the system into operation, NMP Unit 1 UFSAR Section VII VII-11 Rev. 25, October 2017 including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system. The CSS system is designed to permit appropriate periodic pressure and functional testing. Pumps are periodically tested for flow, developed pressure and automatic initiation. Containment spray injection valves are normally open and are not required to operate. The testing program demonstrates, under simulated conditions, that pump sets can be relied upon to function as they are designed to operate under accident conditions. Periodic spraying of water into the containment is not practical. Therefore, water is recycled back to the suppression pool during tests. Following maintenance that could result in nozzle blockage, a test shall be performed on the spray nozzles. Testing of emergency power sources for containment cooling is periodically performed. The power systems are tested for automatic pickup of load required for the LOCA. Criterion 44 Cooling Water A system to transfer heat from structures, systems, and components important to safety to an ultimate heat sink (UHS) shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available), and for offsite electric power system operation (assuming onsite power is not available), the system safety function can be accomplished assuming a single failure. Heat removal from containment following a LOCA, and transferring that energy to the UHS, is achieved and assured through the use of redundant pump trains drawing suction on the suppression pool and removing heat through a heat exchanger supplied by the raw water pumps. The system is designed and suitably sized to maintain the torus below the NPSH temperature limits of core spray and containment spray. NMP Unit 1 UFSAR Section VII VII-12 Rev. 25, October 2017 2.0 Design Bases 2.1 Design Basis Functional Requirements The CSS system shall perform the following functions important to safety in order to prevent containment pressure and temperature from exceeding its design values for reactor coolant system (RCS) leaks up to and including the DBA, double-ended break of a reactor coolant recirculation line: 1. Functional Requirement - Remove energy from the drywell and torus following vessel leaks, up to and including a LOCA, to reduce containment temperature and pressure and maintain them below containment design pressure and temperature limits. Basis - A means of removing energy from containment following a LOCA and of transferring energy to the UHS is required by GDC 38 and GDC 44. The CSS system provides the primary means of energy removal from containment after a LOCA. 2. Functional Requirement - Ensure the torus water temperature does not exceed that required to satisfy containment spray and core spray NPSH requirements. Basis - Inadequate NPSH can limit the containment spray and containment raw water pump performance and reliability. Without adequate NPSH, the ability of the system to remove energy from containment may be diminished. 3. Functional Requirement - Provide the capability to isolate CSS system piping that penetrates the containment boundary. Basis - Unit 1 did not commit to providing isolation valves in the CSS system as would be required to satisfy GDC 56. Containment spray was originally designed as an extension of primary containment. However, Unit 1 has committed to maintaining a water seal in lieu of leak rate testing of the isolation valves. 4. Functional Requirement - The CSS system piping must provide an essentially leak-tight barrier against the NMP Unit 1 UFSAR Section VII VII-13 Rev. 25, October 2017 uncontrolled release of radioactivity to the environment. Basis - The CSS system was originally designed as an extension of primary containment. As such, the containment spray piping must satisfy the intent of GDC 16 and provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment. 5. Functional Requirement - Remove airborne fission products from the drywell atmosphere following a LOCA which results in significant fuel damage, to limit fission product releases from containment leakage paths. Basis - The LOCA radiological analysis implementing the alternative source term (AST) methodology described in Regulatory Guide (RG) 1.183 credits airborne fission product removal by the CSS system. The AST analysis is described in Section XV. 2.2 Controlling Parameters To meet the design requirements of Section VII-B-2.1, the CSS system must be capable of meeting the following operational requirements:

• CSS pump flow through the drywell sparger nozzles must be 3300 gpm.
• CSS pump flow through the torus sparger nozzles must be 300 gpm.
• CSS drywell and torus sparger spray droplet size must be 1000 microns.
• CSS pump flow in the torus cooling mode must be 2800 gpm.
• CSS shell side heat exchanger flow must be 3600 gpm (during containment spray).
• CSS pump available NPSH must be 34.2 ft for the most restrictive case (least NPSH margin) in which two pumps are operating through separate strainer assemblies at a flow rate of 3759 gpm.

NMP Unit 1 UFSAR Section VII VII-14 Rev. 25, October 2017

• CSS raw water pump flow, through the heat exchanger tube side, must be 3000 gpm.
• CSS raw water pump available NPSH must be 31 ft.
• CSS drywell and torus sparger nozzle pressure must be 30 psi above containment pressure for a sufficient number of nozzles to achieve minimum required flows.
• CSS spray header pressure must be 110 percent of containment pressure or 38.5 psig.
• CSS heat exchangers must be capable of removing at least 120 million Btu/hr, with two containment spray pumps operating and a spray water temperature reduction from 140°F to 100°F. 3.0 System Design 3.1 System Function The CSS system is an engineered safeguards system designed to prevent overheating and overpressurization of the containment, reduce drywell airborne fission product concentrations, and control the pressure suppression chamber water temperature following a design basis LOCA. The system is designed to provide heat removal capabilities for vessel leaks up to and including the DBA, the double-ended break of a reactor recirculation line, without core spray system operation. 3.2 System Design Description As shown on Figure VII-3, the CSS system is designed with two redundant loops. The primary loop (Loop 11) provides water to the primary or inner drywell sparger and to the torus sparger. The secondary loop (Loop 12) provides water to the secondary or outer drywell sparger and to the torus sparger. The torus sparger is common to both loops. Each of the two loops are cross-connected through the test return lines such that each of the loops can provide flow to both the primary and secondary spargers. Each loop includes two redundant trains and consists of two suction headers, two containment spray pumps, two heat exchangers and the associated containment spray raw water pumps, a common test return line, and associated piping and control valves. All pumps in a loop are powered from the same emergency NMP Unit 1 UFSAR Section VII VII-15 Rev. 25, October 2017 power bus. Each loop is electrically independent from the other loop. The CSS system is normally in standby. Containment spray pump operation is automatically initiated by two RPS signals--high drywell pressure and low-low reactor water level. Automatic initiation of the containment spray pumps occurs following the core spray pumps and core spray topping pumps initiation. Upon receipt of an actuating signal, the four containment spray pumps are sequentially started when powered from either the reserve Station service or the diesel generators. Upon containment spray pump initiation, self-actuating check valves open to allow containment spray water to flow through the system. The containment spray raw water pumps must be manually initiated following automatic initiation of the containment spray pumps. A 15-min delay can be tolerated in starting a raw water pump since it provides lake water to a containment spray heat exchanger for the purpose of long-term cooling of the torus water. Each pump takes suction from the torus through individual suction lines. The water in each suction line flows from the torus through a suction strainer assembly. Two strainers comprise each of two suction strainer assemblies. When two pumps, either 112 and 122, or 111 and 121, are operated, they will take suction from the same suction strainer assembly. The discharge from each pump passes through the shell side of a heat exchanger where it is cooled prior to being distributed to the drywell and torus spray headers. The spraying of the water in the containment increases the heat removal rate, thereby decreasing containment temperature and pressure. The spray headers inside the drywell and torus are arranged to distribute water as uniformly as possible throughout the free volume. The direction of spray from the nozzles is arranged to minimize impact on equipment and allow as much free-fall as possible to maximize steam condensation. In addition, flow from the containment spray pump discharge can be directed to the torus via a 6-in test return line that provides suppression pool cooling. Each of the containment spray heat exchangers is supplied cooling water from a dedicated containment spray raw water pump. Each containment spray raw water pump takes suction from the condenser circulating water intake tunnel. The pump discharge passes through a duplex strainer prior to entering the tube side of the containment spray heat exchanger. After passing through NMP Unit 1 UFSAR Section VII VII-16 Rev. 25, October 2017 the heat exchanger and cooling the suppression pool water, the raw water is released to the discharge manifold. In the event of a total loss of the containment spray primary water source (suppression chamber water below the containment spray pump suction level), raw water pumps 112 and 121 can be aligned to supply the containment spray spargers to provide an alternate source of containment cooling. Likewise, raw water pumps 111 and 122 can be aligned to supply the core spray system. 3.3 System Design The CSS system was originally designed to operate with Loop 11 and Loop 12 flow paths in the drywell as totally independent redundant systems. However, in order to satisfy 10CFR50 Appendix J, paragraph III.C.3(b) requirements, the current standby configuration of the system provides flow to both primary and secondary spargers, with two pumps including either train 111 or 122 in operation, to form a water seal. This is accomplished by cross-connecting the two trains via the test return line. In this configuration, sufficient header pressures and flows must be provided to meet the Appendix J requirements. Water pressure in the containment spray piping at the drywell and torus spray line penetrations must be at least 110 percent of containment pressure whenever the CSS system is required to spray in the drywell and torus. The calculated peak containment internal pressure is 35 psig, hence, 38.5 psig (3.5 psid) must be maintained in the header to ensure a water seal. In order to meet the water seal configuration, at least two containment spray pumps are required to operate. Calculation shows that in the limiting case when drywell pressure is 35 psig, the following differential pressures between the sparger header and the drywell would be achieved.

NMP Unit 1 UFSAR Section VII VII-17 Rev. 25, October 2017 Pumps Operating Minimum Drywell Sparger Header Pressure (psid) Minimum Torus Sparger Header Pressure (psid) Primary Loop Secondary Loop 2 Primary (111 & 112) Separate Strainer Assemblies 2 Secondary (122 & 121) Separate Strainer Assemblies 50 18 16 54 41 43 Spray heat removal effectiveness is a function of droplet size and spray flow rate. The design spray flow requirement to the drywell and torus to remove all decay heat and chemical energy from a 70-percent metal-water reaction was calculated to be 3600 gpm (3300 gpm to the drywell and 300 gpm to the torus). This is the containment spray pump design point. Each pump is rated for 3000 gpm at 375 ft (162 psig). Calculation shows that each pump can deliver sufficient flow under DBA LOCA conditions to meet the minimum required flow. The following shows that the required flows can be met. Calculated Flow Rate (gpm) Required Flow Rate (gpm) (Table XV-32a) One-Pump Operation (121) Containment Spray Mode One-Pump Operation (111) Torus Cooling Mode One-Pump Operation (112) Torus Cooling Mode One-Pump Operation (121) Torus Cooling Mode One-Pump Operation (122) Torus Cooling Mode Two-Pump Operation (111 & 112) 3650 2850 2822 2953 2816 3605 (Pump 111) 3600 2800 2800 2800 2800 3000 per pump NMP Unit 1 UFSAR Section VII VII-18 Rev. 25, October 2017 Calculated Flow Rate (gpm) Required Flow Rate (gpm) (Table XV-32a) Containment Spray Mode Separate Strainer Assemblies Two-Pump Operation (121 & 122) Containment Spray Mode Separate Strainer Assemblies 3425 (Pump 112) 3438 (Pump 121) 3553 (Pump 122) 3000 per pump To determine the expected droplet size distribution, tests were performed on spray nozzles similar to those used in the CSS system. The test results show that a droplet size <1000 microns is acceptable. A nozzle pressure differential 30 psi is required to achieve the desired droplet size. This requirement is met for the minimum flow required at DBA LOCA conditions as shown above. Each operating loop, with a total pump flow rate of 6000 gpm, has a heat removal capacity of 120 million Btu/hr with a spray water temperature reduction from 140°F to 100°F as it passes through the heat exchangers. This is the original design basis heat exchanger sizing point. The design basis heat removal requirements associated with a maximum containment spray raw water temperature of 84°F is provided in Section XV-C-5.3. This analysis results in a peak suppression pool temperature of 165°F. The corresponding containment spray heat exchanger K-value in the spray mode is 256 Btu/sec-°F, and in the torus cooling mode is 241 Btu/sec-°F. Therefore, for the conditions specified, the heat exchangers are capable of removing enough heat such that torus and drywell temperature and pressure limits are not exceeded. The raw water side of the heat exchanger is maintained at a greater pressure than the containment water side to avoid contamination of the environment in the event of a tube leak. Radiation detection alarms are located on the heat exchanger raw water discharge. The containment spray raw water cooling system is considered operable when the pressure on the raw water side of the heat exchanger is greater than 141 psig and when the flow is greater than 3000 gpm. The containment side of the heat exchanger will operate at less than 117 psig, including containment pressure under accident conditions. NRC Bulletin 96-03 requested implementation of appropriate procedural measures and plant modifications to minimize the NMP Unit 1 UFSAR Section VII VII-19 Rev. 25, October 2017 potential for clogging of the ECCS suction strainers as a result of debris accumulation from debris generated during a postulated LOCA. Large-capacity ECCS suction strainers have been designed and installed to account for the worst-case generation, transport, and accumulation of post-LOCA debris to assure a sufficient available NPSH to the pumps. Containment spray pumps NPSH was calculated for the bounding conditions as set forth by Regulatory Guide (RG) 1.1. The conditions that result in the minimum NPSH margin (available NPSH minus required NPSH) was determined to be two operating containment spray pumps drawing suction from the same strainer assembly, less than 15 min after the onset of the LOCA, when torus temperature is 146°F, and containment pressure is 0 psig. These conditions provide the minimum NPSH margin of 1.0 ft, assuring that adequate NPSH will exist for all operating conditions and operating modes. Containment spray raw water pumps NPSH has been calculated for the limiting conditions of maximum lake water temperature of 83°F and minimum lake water level of 238.5 ft at the screenhouse. The calculated available NPSH for all four pumps at the design flow of 3000 gpm exceeds the required NPSH. Vendor-supplied pump curves indicate that the pumps can operate up to 3600 gpm without exceeding the available NPSH. The torus cooling mode of operation is used for long-term cooling of the suppression pool. In this mode, all containment spray pump flow is directed to the torus through the test return line. When operating in accordance with the EOPs with only one pump operating, the flow through the heat exchanger is at least 2800 gpm for all pumps. The piping, heat exchangers, and other equipment are designed for containment pressure and pump shutoff head. The following table summarizes the design pressures of the system's major components. Equipment Design Pressure (psig) Containment spray piping outside drywell Containment spray piping inside drywell Heat exchanger tube side Heat exchanger shell side Raw water piping 270 235 270 220 300 NMP Unit 1 UFSAR Section VII VII-20 Rev. 25, October 2017 3.4 Codes and Standards The piping and valves are designed in accordance with ASA B31.1-1955 Piping Code with certain requirements of the ASME Code Section III-B-1965. The containment spray pump casings and containment spray strainer housings are designed in accordance with ASME Code Section III-1965. The raw water pump casings are designed in accordance with ASME Code Section III-1965, and raw water pump strainer housings are designed in accordance with ASME Code Section VIII. The heat exchangers were designed, fabricated, tested and certified to the 1980 Edition of the ASME Boiler and Pressure Vessel Code Section III (Class 2 for the shell side and Class 3 for the tube side). Valves 80-114 and 80-115 were designed and fabricated in accordance with ASME Code Section III-1977, and tested to ANSI B16.104-76 Class VI for seat leak test. The Mark I Program suction piping is designed in accordance with ASME Code Section III-1977. All electrical power and distribution is designed in accordance with IEEE-279, IEEE-308, IEEE-323, IEEE-336 and IEEE-344 (all 1971 Editions). No single active failure of any component can prevent the system from fulfilling its design function. 3.5 System Instrumentation Each CSS system loop has separate and independent pressure and flow indicators. In addition, temperature indications of containment spray water in and out of the containment spray heat exchangers are provided. High temperature alarms are provided for heat exchangers outlet temperatures. Isolation valve control switches and position indicator lights are provided on the main control panel for each isolation valve. Indication of isolation valve position is repeated on the isolation valve mimic on the main control room panel. Each containment spray pump has its own control switch with indicating lights and a motor ammeter on the control room panel. Pressure switches on each loop header are outside the drywell and will indicate low containment spray header pressure if the system has been called upon to operate. For each line carrying cooling water to the heat exchangers, flow indication is provided on the main control room panel. Each containment spray raw water pump discharge strainer is monitored for plugging by differential pressure switches. Heat exchanger cooling water effluent is monitored for high radiation before discharging to the tunnel. These alarms are sounded on the control room annunciators. NMP Unit 1 UFSAR Section VII VII-21 Rev. 25, October 2017 All sensing instrumentation is in areas accessible during Station operation and is provided with suitable valving for in-place testing at any time. 3.6 System Design Features The CSS system is designed to provide a high degree of reliability in meeting the design functional requirements. The specific features of the system that assist in achieving this reliability are:

• Each loop is designed with heat removal capacity well in excess of the expected maximum.
• The two loops are sufficiently separated to minimize the possibility of coincident active failures.
• The system is designed to meet any credible seismic force as discussed in Section XVI-C.
• Automatic initiation of all four pumps of the CSS system assures that the containment will not be overpressurized.
• Electric power for the system is available from Station reserve power supplies or from either of two emergency diesel generators.
• The raw water side of the heat exchangers is operated at a higher pressure than the containment water side to prevent out-leakage. Radiation monitors are installed to detect such leakage if it should occur when the raw water pumps are not operating.
• The delay between the time of the accident and full spray operation is less than a minute. This includes signal time for pump start, time required to get pumps up to speed, and diesel generator starting time. Even assuming no core spray and the maximum metal-water reaction, this delay could be as much as 15 min without loss of containment integrity.
• Low pressure alarms are provided in the containment spray piping outside the drywell to signal the Operator if spray water is not reaching the upper nozzles due to a restriction or line break.

NMP Unit 1 UFSAR Section VII VII-22 Rev. 25, October 2017

• Remotely-operated valves are located on the containment spray bypass line to the waste disposal building to provide isolation of a potential pathway for the transfer of radioactivity, should a LOCA occur during suppression chamber pumpdown.
• Raw lake water can be supplied to the containment spray nozzles as an alternate source of containment cooling. 4.0 Design Performance Evaluation The performance of the CSS system is determined through application of the 10CFR50 Appendix K evaluation. The SAFER/GESTR-LOCA Analysis was used to evaluate the ECCS (including CSS system) performance during a postulated LOCA. The details of the analysis are discussed in Chapter XV. 4.1 System Performance Analyses Analysis has been performed which supports the adequacy of the CSS system in maintaining containment pressure and temperature below the design values following a design basis LOCA. The Section XV-C-5.3 design basis reconstitution suppression chamber heatup analysis verifies that the containment design basis heat removal requirements are satisfied at the maximum containment spray raw water (lake water) temperature of 84°F. Each of the two containment spray loops was originally sized to remove all decay heat and chemical energy from a 70-percent metal-water reaction. With a maximum possible reaction of 27 percent, the analysis shows that more than sufficient heat removal capacity exists in the system. This analysis requires the CSS system to satisfy the analysis input assumptions discussed in Section XV-C-5.3.2. To determine proper distribution of containment spray through the nozzles, testing was performed on a sample spray nozzle of the size and type used in containment spray. Water was run through the nozzle at various pressures from 10 psig to 100 psig, and spray pattern and spray particle fineness was observed. Pressure drops of 80 psig and 30 psig represent the original system configuration pressure conditions for two-pump operation and one-pump operation, respectively. The particle sizes for the two-pump operation are in the range of 10 to 400 microns. For one-pump operation, particle sizes range from 500 to 1000 microns.

NMP Unit 1 UFSAR Section VII VII-23 Rev. 25, October 2017 The CSS system design flow, spray distribution/droplet size, and fall heights were used to determine the airborne fission product removal rate for implementation of the AST methodology described in Section XV. 4.2 System Response After an initiation signal is received, there is a time delay of 20 sec to allow the core spray and core spray topping pumps to start. At the 25-sec mark, containment spray pumps 111 and 121 will receive a start signal, and at 30 sec, containment spray pumps 112 and 122 will receive their start signal. If the core spray and core spray topping pumps do not start, a set of backup timer contacts will start the containment spray start sequence in 50 sec to allow the core spray starting logic to be initiated a second time. This will cause pumps 111 and 121 to start at 55 sec, and pumps 112 and 122 to start in 60 sec. This interlock, delaying the starting of the containment spray pumps, is provided to avoid overloading of the diesel generators. 4.3 Interdependency With Other Engineered Safeguards Systems The CSS system is used in conjunction with the core spray system described in Section VII-A. The core spray system removes heat from the core in the event of a LOCA. In the heat removal process, the core spray water is converted to steam, which is then released to the containment. The containment sprays condense the steam in the drywell and remove heat from the containment vessels through heat exchangers. The raw water pumps are interconnected with the core spray system and the containment spray loops to provide an emergency source of water. Raw water pump 112 can supply water to containment spray train 122, and raw water pump 121 can supply water to containment spray train 111. The motor-operated valves between raw water and containment spray water are interlocked with the heat exchanger raw water discharge valves. If one valve is open, the other must be closed. In addition, raw water pump 111 is connected to core spray pump train 11 and raw water pump 122 is connected to core spray pump train 12. The air-operated valves located on the connection between the two systems are also interlocked with the raw water discharge valves. The following systems must be in operation to support the CSS system: NMP Unit 1 UFSAR Section VII VII-24 Rev. 25, October 2017

• Instrument air must be operational to permit operation of the containment spray inlet isolation valves and bypass blocking valves.
• 4.16-kV and 600-V ac power distribution systems are required to provide power to the containment spray pumps, raw water pumps, and isolation valves.
• The RPS system is required to provide automatic initiation signals to the containment spray pumps and waste disposal isolation valves.
• The process radiation monitoring system must be operational to alert Operators of leakage of contamination into the raw water system due to heat exchanger leaks. 5.0 System Operation 5.1 Limiting Conditions for Operation The limiting conditions for operation (LCO) pertaining to the CSS system are listed in Section 3.3.7 of the Unit 1 Technical Specifications. Other LCOs associated with generic equipment and programs are also applicable and are listed in other sections. The intent of the LCOs is to ensure that both loops of the system are operable when fuel is in the vessel and the reactor coolant temperature is greater than 215°F. One containment spray loop will provide the required containment cooling, airborne fission product removal, and pressure reduction for the DBA. However, to provide sufficient redundancy to satisfy the single failure criterion, both loops of the CSS system are required to be operable. If a redundant component in one loop of containment spray or its associated raw water loop becomes inoperable, operation may continue provided the component is returned to an operable condition within 15 days. If a redundant component in both containment spray loops or their associated raw water loops becomes inoperable, operation may continue provided the component is returned to service within 7 days. In both cases, additional surveillance requirements are imposed. If a containment spray loop or its associated raw water loop becomes inoperable and all the components of the other loop are operable, the reactor may remain in operation for a period not to exceed 7 days.

NMP Unit 1 UFSAR Section VII VII-25 Rev. 25, October 2017 If the LCOs are not met, then a normal orderly shutdown shall be initiated within 1 hr and the reactor shall be placed in cold shutdown within 10 hr. 6.0 Tests and Inspection To ensure that the performance of the CSS system continues to meet the design requirements, the following surveillance tests and inservice inspections requirements must be satisfied.

• ASME Section XI inservice examination of components
• ASME OM Code inservice testing of pumps and valves
• ASME Section XI system pressure tests
• Appendix J leak rate testing
• System operability surveillance tests Several programs have been established to meet the requirements of the ASME Code and Appendix J. These include: 1) NMP1 ISI Program Plan, 2) Inservice Pressure Testing Program Plan, 3) Pump and Valve Inservice Testing Program Plan, and 4) Appendix J Testing Program Plan. The following CSS system tests, inspections, and surveillances are conducted to meet the requirements.
• Containment Spray System Quarterly Operability Test - verifies valve, pump and total system operability and verifies operation of valve limit switches and solenoid-operated valves
• Containment Spray Header and Nozzle Air Flow Test - verifies header, header check valve, and nozzle operability
• Containment Spray System Suction Valve Operability Test - verifies valve operability
• Containment Spray Valve Remote Position Indicator Verification - verifies operability of indicators
• Containment Spray Pressure Test - verifies integrity of the system by VT-2 visual examination NMP Unit 1 UFSAR Section VII VII-26 Rev. 25, October 2017
• Containment Spray Raw Water Pressure Test - verifies integrity of the system by VT-2 visual examination
• Containment Spray Raw Water System Intertie Valve Operability Test - verifies the operability of the containment spray/core spray intertie check valves Testing of the initiating instrumentation and controls portion of the system is discussed in Section VIII. The emergency power system, which supplies electrical power to containment spray in the event that offsite power is unavailable, is tested as described in Section IX. Visual inspections of all system components located outside the drywell can be made at any time during power operation. Components inside the drywell can be visually inspected only during periods of access to the drywell. C. LIQUID POISON INJECTION SYSTEM 1.0 Design Bases The liquid poison injection system is provided to bring the reactor to a cold shutdown condition at any time in core life independent of the control rod system capabilities. Cycle-specific analysis results are contained in the SRLR(7). The primary requirement imposed on the liquid poison injection system is to shut down the reactor from a full-power operating condition, assuming complete failure of the withdrawn control rods to respond to an insertion signal. Injection of liquid poison is also required following a large break LOCA to maintain the suppression pool water pH > 7.0 in support of the AST methodology. For the design rating of 1850 MWt, a concentration of 109.8 ppm of boron-10 isotope (equivalent to 600 ppm of natural boron) is required in the reactor to meet the reactor shutdown requirement. However, an additional 25-percent margin is included in the calculation of required liquid poison tank concentrations to allow for nonuniform mixing of the liquid poison as it is injected into the reactor. The same tank concentration level has been determined to adequately satisfy the AST support function for controlling pH above 7.0. The rate of reactivity compensation provided by the liquid poison injection system is designed to exceed the rate of reactivity gain associated with reactor cooldown from the full-power condition. The liquid poison system is not intended NMP Unit 1 UFSAR Section VII VII-27 Rev. 25, October 2017 to be capable of producing as rapid a shutdown as is produced by scramming the control rods, and should not be construed as a scram backup. Following a large break LOCA, initiation of the liquid poison system within 1.5 hr after the potential for significant fuel failure has been identified will ensure that the suppression pool pH is controlled for at least 30 days. The liquid poison injection system is actuated only by remote manual action from the control room, hence a deliberate action. The liquid poison injection system can be powered from the diesel generators and, therefore, will be operable in the event of a loss of normal and reserve ac power. The liquid poison system is required to function for a maximum of 3 hr following pipe break events (accidents) that produce harsh environmental conditions. Accordingly, EQ in accordance with 10CFR50.49 for the 3-hr post-LOCA mission time has been demonstrated for the electrical components important to safety that comprise the liquid poison system. EQ in accordance with 10CFR50.49 is not required for anticipated transients without scram (ATWS) that may produce harsh environmental conditions inside containment, but not in the reactor building where the electrical components are located. All portions of the system are designed for earthquake loads of 0.3g horizontal and 0.1g vertical. 2.0 System Design The liquid poison injection system, shown on Figure VII-6, consists of an ambient pressure tank with immersion heater for low-temperature sodium pentaborate solution storage, two high-pressure positive displacement pumps for injecting the solution into the reactor core, two explosive-actuated shear plug valves for isolating the liquid poison from the reactor until required, an in-vessel sparger ring, a test tank, two isolation check valves, a buffer system and additional valves, piping and associated instrumentation. The liquid poison is stored in a 4080-gal tank which is designed for atmospheric pressure. This tank is complete with top cover, hatch with lid for adding chemicals, immersion-type electric heater, instrument connections, and nozzles for outlet, recirculation, overflow, air sparger and drain. The tank outlet nozzle is outfitted with a strainer, which extends above the tank bottom, to prevent solid particles from being discharged to NMP Unit 1 UFSAR Section VII VII-28 Rev. 25, October 2017 the pump suction. The air sparger, which is used for mixing the solution for each initial batch, has air holes directed toward the bottom of the tank for sweeping the deposit there. The top cover and hatch lid are designed so that the solution, when agitated by the air sparger, will not spill over. The neutron absorber in the sodium pentaborate liquid poison solution is the boron-10 isotope. The relationship between liquid poison solution concentration and boron-10 enrichment is contained in the equivalency equation. The equation is: Where: C = Sodium pentaborate solution concentration (wt %) M = Mass of water in reactor vessel and recirculation piping at hot rated conditions (501500 lb) Q = Liquid poison pump flow rate (30 gpm nominal) E = Boron-10 enrichment (Atom %) The saturation temperature varies with solution concentrations of sodium pentaborate as shown on Figure VII-7. This saturation curve has 5°F margin above the actual saturation temperature to prevent precipitation of sodium pentaborate while in storage. The liquid poison tank contains a minimum volume of 1325 gal of sodium pentaborate solution whose (boron-10) enrichment and concentration conform to the equivalency equation. To compensate for evaporation which could lead to precipitation, the storage tank was oversized. The nominal tank capacity of 4080 gal allows additional water to be added to the solution as a safety margin against evaporation losses. Temperature and liquid level alarms for the storage tank are annunciated in the control room. The 50-kW, 550-V three-phase immersion heater is automatically controlled by a temperature indicator controller. High- and low-temperature annunciators are provided to assure that the solution is above saturation temperature. Pump test results indicate adequate NPSH is available at solution temperatures up NMP Unit 1 UFSAR Section VII VII-29 Rev. 25, October 2017 to and including 105°F. Solution temperatures up to 130°F have been analyzed and also provide adequate NPSH. To increase the rate of sodium pentaborate solution in water, a manual override on the temperature controller permits heater operation for 150°F solution temperature. This manual override may render the system inoperable. An indicator lamp is provided to denote when the heater element is shorted to the solution. Should the immersion heater fail during Station operation, no action need be taken. Normally, the building heating system will maintain the required tank temperature. The immersion heater is used to supply the endothermic heat required during solution mixing and only incidentally to maintain solution temperature. If a failure of the building heating system occurs simultaneously with a failure of the immersion heater, the ambient temperature in the liquid poison system area will decrease very slowly due to its large mass and its interior building location. Therefore, there will be ample time to provide temporary heating. The sodium pentaborate solution is delivered to the reactor by one of two 30-gpm, positive displacement pumps, with a design discharge pressure of 1670 psig. The pumps and piping are protected from overpressure by two relief valves which discharge back to the poison storage tank. The relief valves are set to open at a pressure between 1455 and 1545 psig. The injection pumps produce a flow rate sufficient to meet the injection requirements for all conditions of reactor operation up to the primary system design pressure of 1250 psig. Two pumps are provided to give complete redundancy. The pumps are specifically designed for standby service to be operated infrequently, only during emergencies and testing. Each operation is for 3 hr maximum. Since the liquid poison injection system is to be operable in the event of loss of normal and reserve ac power, one pump is connected to PB 102 and the other to PB 103. These boards are powered from the diesel generators in the event of failure of their normal supply as described in Section IX, Electrical Systems. A radiant heat shield is installed between the two liquid poison pumps to prevent fire damage to the redundant pump in the event of a fire in either pump. The explosive valves are double squib-actuated shear plug valves. A low-current electrical monitoring system gives visible NMP Unit 1 UFSAR Section VII VII-30 Rev. 25, October 2017 (pilot light) and analog (ammeter) indication of circuit continuity through both firing squibs in each valve. Operation of one valve provides sufficient flow passage to meet the required flow rate. Two valves are provided to give complete redundancy. The firing reliability of these explosive valves is in excess of 99.99 percent. The approximate firing current is 2 amps and the operating time at 2 amps is a nominal 0.002 sec. The products of the explosion are completely contained. The buffer system is composed of gas-charged diaphragm accumulators of the capacity required to absorb fluid pulsation initiated by the positive displacement pumps. Each is located as close as possible to its respective pump discharge. Each pump loop has an accumulator. Containment isolation is provided by two check valves in the liquid poison pipe, one check valve just outside the drywell penetration and the other check valve inside the drywell. An additional check valve is installed downstream of each relief valve connection. The purpose of each check valve is to prevent flow through an assumed defective relief valve of the idle pump loop while the second loop is in operation. This ensures that the capacity of the second pump remains unaffected. The liquid poison sparger in the reactor pressure vessel (RPV) is a 1-in stainless steel pipe which is fastened to the inside of the vessel shroud below the core support plate. This 360-deg sparger has ten 1/4-in drilled holes which are distributed equally around the sparger and which spray toward the bottom of the vessel. The liquid poison is thereby mixed with the reactor recirculating water as it enters the reactor fuel assemblies, assuring as uniform a mixture of poison as practical. During injection following a LOCA, the solution is mixed in the vessel bottom head with core spray water flowing through the reactor core and out the break. A test tank and demineralized water supply are an integral part of the system to facilitate system testing and flushing. All piping in the system has been designed in accordance with ASA B31.1-1955 Piping Code. Tanks are constructed in accordance with API 650. The pressure-bearing parts of the pumps are built in accordance with ASME Code Section III, Class C-1965.

NMP Unit 1 UFSAR Section VII VII-31 Rev. 25, October 2017 Actuation of the liquid poison system is manually initiated from the control room, assuring that poison injection is caused by a deliberate act. 2.1 Operator Assessment The Operator can assess operation of the liquid poison system by means of pressure indication and pump motor ammeters on the main control room panels. Each explosive valve has a low-current electrical monitoring system with an ammeter and lights on the main control room panel. The ammeter and lights provide indication of circuit continuity through both firing squibs in each valve; the ammeter and lights ensure firing readiness. When fired, the circuit is broken, the ammeter reads 0, the indicating lights go out, and the control room annunciator alarms. The pressure transmitter is in an accessible location and is provided with suitable valving so that it can be tested at any time. The liquid poison tank is provided with control room level indication in addition to alarms for high- and low-level and high- and low-solution temperature. 3.0 Design Evaluation The liquid poison system is designed to provide the capability to bring the reactor from full design rating (1850 MWt) to a cold, xenon-free shutdown condition assuming none of the control rods can be inserted, and to buffer the suppression pool water following a large break LOCA. To meet the shutdown objective, the system is designed to inject a quantity of boron which produces a concentration of at least 109.8 ppm of boron-10 isotope in the reactor core. This concentration will bring the reactor from full design rating (1850 MWt) to a subcritical condition considering the combined effects of the control rods, coolant voids, temperature change, fuel doppler, xenon, and samarium. The same quantity of boron will maintain the suppression pool pH 7.0 for 30 days following a LOCA that results in significant fuel damage. Cycle-specific analysis results are contained in the SRLR(7). The minimum required tank storage volume and conformance of the liquid poison solution concentration and boron-10 isotope enrichment assures that the expected liquid poison solution will provide the required 109.8 ppm of boron-10 isotope to the NMP Unit 1 UFSAR Section VII VII-32 Rev. 25, October 2017 reactor core or the necessary buffering solution to the suppression pool. The liquid poison storage tank volume concentration requirements assure that the above requirements for boron solution insertion are met with one 30-gpm liquid poison pump. Normal level is maintained between 1400 and 1700 gal. The quantity of boron-10 isotope required to be stored in solution includes an additional 25 percent margin beyond the amount needed to shut down the reactor to allow for any unexpected non-uniform mixing. The minimum tank volume requirements include consideration for 197 gal of solution which is contained below the point where the pump takes suction from the tank and, therefore, cannot be inserted into the reactor. The solution saturation temperature varies with the concentration of sodium pentaborate. Solution temperature is maintained by Technical Specification at least 5°F above saturation temperature to guard against precipitation. Temperature and liquid level alarms for the system are annunciated in the control room. Equipment and piping are designed to withstand the most severe conditions of loads including the design earthquake. Nozzles leading into the reactor vessel have been designed taking into account possible vessel movement due to an earthquake. Availability of emergency diesel generator power to both of the injection pumps assures operability of the system if required during a loss of normal and reserve ac power. 4.0 Tests and Inspections The system has been designed to permit periodic testing, maintenance, and operation of the injection pumps and appropriate valves. The pumps and valves will be tested periodically to ensure operability. Monthly pump tests are performed during Station operation either with demineralized water recirculated to the test tank, or with the solution recirculated to the poison tank. The isolation valves may be tested only during shutdown. For explosive valve tests, the valves are dismantled and inspected. The charges are removed and replaced with new charges periodically and the old charges are test fired to establish a rational charge replacement frequency. NMP Unit 1 UFSAR Section VII VII-33 Rev. 25, October 2017 A demineralized water purge system is provided so that the remaining portion of the system may be tested by pumping demineralized water through the distribution system and into the reactor vessel once each operating cycle. Boron concentration and boron-10 enrichment of the solution will be periodically determined by analysis. The temperature of the solution will be monitored and annunciated in the control room to assure that the solution is above its saturation temperature. A continuity check of the firing circuit on the explosive valves is provided by pilot lights in the control room. The functional test and other surveillance of components, along with the monitoring instrumentation, gives a high reliability for liquid poison injection system operability. 5.0 Alternate Boron Injection In the event the liquid poison system is not available, the EOPs list alternate methods for injecting boron into the reactor. One method, referred to as the alternate boron injection system, provides for a portable pneumatic hydro pump connection between the liquid poison tank overflow/drain line and the liquid poison injection line drain valves, as shown on Figure VII-6. Borated water is then suctioned from the liquid poison tank through the hydro pump and discharged into the existing liquid poison injection line. The air supply required for the pneumatic hydro pump can be provided from a 1-in connection to the house service air, or from the instrument air system if house air is not available. The portable hydro pump has a design flow rate of 7.5 gpm at 1460 psig. The design pressure and flow rate of the hydro pump are sufficient to provide flow to the vessel under a worst-case vessel pressure of 1339 psig using an enriched boron solution. The hoses (suction, discharge and air hose) and the portable hydro pump for the alternate boron injection system are stored in the vicinity of the 55-gal drum in the reactor building. The hoses are in a locked compartment to assure their availability. The alternate boron injection system is nonsafety related and nonseismic, as the additional hoses, pump, valves, and hose connections do not perform a safety-related function and are downstream of the safety-related portions of the liquid poison system. NMP Unit 1 UFSAR Section VII VII-34 Rev. 25, October 2017 The other alternate boron injection method, using the reactor cleanup system, provides for filling the cleanup filter with a boron solution and injecting the solution into the vessel by placing the filter in service. Neither alternate system is expected to be available for injection following a LOCA, since harsh environments in the reactor building will prevent the required operator access. D. CONTROL ROD VELOCITY LIMITER 1.0 Design Bases The control rod velocity limiter, in conjunction with the rod worth minimizer (RWM), is provided to limit any accidental reactivity addition to rates for which the resulting excursion would not rupture the pressure vessel or impair operation of any safeguards equipment. The worst reactivity addition occurs during the control rod drop accident (CRDA) (Section XV), the consequences of which are reactivity rate dependent. The control rod velocity limiter is an engineered safeguard that was originally designed to limit the free-fall drop velocity of the control rod to 5 ft/sec or less and, thus, limit the rate of reactivity addition. Subsequent testing and analysis demonstrated a maximum rod drop velocity of 3.11 ft/sec for use in CRDA analyses(6). The CRDA can only happen in the event of simultaneous procedural violations and equipment malfunctions, a separation or mechanical failure in the drive line, sticking or binding of the control rod, the withdrawal of the detached control rod drive (CRD) mechanism, and then the release of the control rod by some unspecified means. The rod velocity limiter is designed to limit the consequences of the drop of the maximum worth control rod without significantly hindering the normal function of the system. The most probable threshold for potential mechanical damage to the reactor core or other primary cooling system components is a peak fuel enthalpy in excess of 425 cal/g. By reducing the velocity of a free-falling rod, and assuring that excessive rod worth patterns are not established, the CRDA will result in peak fuel enthalpy values below the design limit of 280 cal/g.(1) 2.0 System Design NMP Unit 1 UFSAR Section VII VII-35 Rev. 25, October 2017 The control rod velocity limiter is an integral part of the bottom of each control rod, as shown on Figure VII-9 (typical). It is designed as a large clearance piston which travels in the control rod guide tube over the entire stroke. The original velocity limiter assembly consists of two conical elements machined from a single 304 stainless steel casting. The lower conical element is at a 15-deg angle relative to the upper conical element, and the two elements are separated with four spacers 90 deg apart. There are no moving parts in the velocity limiter. The rod velocity limiter provides a streamlined profile in the scram (upward) direction and a nonstreamlined profile in the dropout (downward) direction. It may be regarded as a nozzle-type limiter since, during its downward motion, a high percentage of the total water directly below the limiter flows up through the center of the limiter body and is ejected radially outward into the limiter guide tube annulus at an oblique angle due to its conical configuration. The rod velocity limiter has a maximum diameter of 9.265 in and is fitted inside the control rod guide tube, which has an inside diameter of 10.420 in. This configuration results in an annulus between the limiter and the guide tube of 0.577 in. The casting terminates at the top in a cruciform section which matches the blade shape. The limiter is welded to the control rod blade at the cruciform to become an integral part of the control rod assembly. The casting is machined to obtain the required outside diameter, roller lug and backseat. Consistent rod velocity limiter performance is assured by machining the outside diameter and maintaining close control of the hydraulically-important casting tolerances. The backseat and coupling are identical to designs presently employed in the locking piston drive. The rollers are Inconel X-750 to prevent galling against the interior of the guide tube and are held in place by pins made of PH13-8Mo alloy. Some blades still have Stellite rollers and Haynes 25 alloy pins. As these blades are replaced, the new low cobalt pin and roller materials will be used. A nominal radial gap of 0.125 in is provided between the rollers and guide tube wall to allow for tolerance variation and misalignment. A coupling release handle is located in the cruciform section immediately below the control rod blade and rod velocity limiter NMP Unit 1 UFSAR Section VII VII-36 Rev. 25, October 2017 weld; raising the handle permits the control rod assembly to be separated from the locking piston drive. For the advanced long life control rod (ALLCR), the D-230, and the Marathon control rod designs, the velocity limiter was redesigned to compensate for the increased weight of the hafnium. Although the ALLCR, D-230, and Marathon control rods are slightly lighter than the original control rod, scram performance is not adversely affected. The improved control rod is further described in Section IV-B.6.1.1. The guide tube in which the rod velocity limiter acts has a backseat on its lower end that rests on the CRD housing. This seat restricts water flowing out of the guide tube during rod withdrawal or during the control rod dropout. The seat also prevents water bypassing the fuel elements during normal operation by restricting any flow into the guide tube. A close fit is provided between the top of the guide tube and the core plate to restrict flow through this joint, which could also bypass the fuel elements. There is normally an external pressure of ~15 psi on the guide tube due to the core pressure drop. The guide tubes are constructed of 10-in schedule 10 304 stainless steel pipe, which is adequate for an external pressure of 100 psi. A 304 stainless steel fuel support casting rests on top of each guide tube. Each of these castings supports four fuel elements and contains four flow passages, one for each element. Each flow passage has an orifice to regulate the flow through the fuel elements. The fuel support flow passages are aligned with the guide tube orifices by means of a core plate pin. 3.0 Design Evaluation 3.1 General The rod velocity limiter must limit the free-fall velocity without significantly reducing the scram time of the rod. The hydraulic drag forces on the rod are approximately proportional to the square of the rod velocity and are essentially negligible during normal withdrawal or insertion functions and during the first part of the scram stroke. However, during the latter part of the scram stroke, the rods reach relatively high velocities; therefore, the drag forces could become appreciable. As an additional design consideration, the rod velocity limiter must not produce unacceptable loadings on the drive mechanism during the scram action. NMP Unit 1 UFSAR Section VII VII-37 Rev. 25, October 2017 Various full-scale tests were performed at cold and at operating conditions to evaluate the effect of the rod velocity limiter on the control rod scram time. These tests included full-scale prototypes using production design components under simulated reactor operating conditions. Results of these tests show the rod velocity limiter will cause an increase in scram time of approximately 0.25 sec for 90-percent insertion of the stroke. Technical Specification scram time criterion for this distance is 5.0 sec, while the measured average scram times range from 2.4 to 3.1 sec. Therefore, the increased scram time due to the rod velocity limiter is not prohibitive and the overall performance satisfies the design criteria. 3.2 Design Sensitivity Numerous model tests were performed to determine the sensitivity of the design to changes in various dimensions (angles, gaps, etc.). Wooden and metal models were fabricated so that separate parameters could be varied while holding the others constant. Approximately 100 drop tests were performed. As a result of these tests, the optimum design was based on the following principles. 1. The upper and lower cones must converge at 45 deg and 30 deg, respectively, from the horizontal and must be separated by about 1 in at their periphery. 2. The upper cone must have an appreciable thickness. 3. The top of the upper cone requires a gentle radius at its outer edge and must terminate with a sharp edge. 4. A long, straight or tapered skirt attached to the lower cone decreases the effectiveness of this design. 5. The contours through the nozzle area need not be streamlined. 6. The relative angle between the cones, the outside diameter of the cones and the top cone outer edge radius are dimensions which must have close tolerance. 3.3 Normal Operation Normal rod insertion and withdrawal operations were performed to detect possible abnormalities which could be attributed to the NMP Unit 1 UFSAR Section VII VII-38 Rev. 25, October 2017 presence of the rod velocity limiter. As can be seen in Table VII-1, the pressure drop across the piston and the travel times are within the test guideline tolerance. 4.0 Tests and Inspections Because the control rod velocity limiter is an integral part of the control rod assembly, it will not require specific retesting after installation. In addition to close surveillance during fabrication of the rod velocity limiter and control rod assembly, random control rod assemblies undergo shop testing which includes rod drop tests. The control rod assemblies were also tested during preoperational tests prior to reactor startup and after initial installation. These preoperational tests confirmed the operation of the individual control rod assemblies for normal operation and scram conditions. E. CONTROL ROD HOUSING SUPPORT 1.0 Design Bases The CRD housing support system is an elastic structure designed to absorb the maximum possible dynamic loading resulting from a complete instantaneous circumferential failure of a single CRD housing and to limit the resulting control rod ejection to less than one "normal drive notch" position (6 in). In the case of the highest credible rod worths, a high-velocity rod ejection could cause a power excursion with fuel enthalpies exceeding the threshold of sudden fuel rupture (425 cal/gm, see Section IV-B.6.2.2). To provide a margin of safety, there must not be a failure of a housing associated with the CRD mechanism which would permit a significant movement of the rod and its drive at high velocities. The criteria for design of the CRD housing support system are as follows: 1. Limit the control rod motion to approximately 3 in following an assumed CRD housing failure. Despite the conservatism in the housing design, it is assumed that a drive housing could experience a failure allowing the control rod to be driven from the core. The failure resulting in the highest forces is an instantaneous circumferential failure of the CRD housing with the reactor vessel at the design pressure of 1250 psig. NMP Unit 1 UFSAR Section VII VII-39 Rev. 25, October 2017 A failure resulting in lower forces for which the CRD housing support is also designed is an instantaneous failure of all the flange bolts or a housing at the design pressure of 1250 psig. The forces are smaller because the pressure would be acting on the inside rather than the outside of the housing (as in a circumferential housing failure). Again, the CRD housing support system would limit the downward motion of the control rod and its drive to approximately 3 in. 2. Provide clearance between the housing support grid plates and the housings to prevent contact due to their respective expansions during plant operation. 3. Provide a system which is easily removable one module at a time to permit access to and removal of the CRD mechanisms, position indicators, and in-core housings for maintenance and inspection. 2.0 System Design The control rod housing supports are illustrated on Figure VII-10. Horizontal beams are placed between the rows of control rod housings and are supported by the reactor support pedestal immediately below the bottom head of the reactor vessel. The beams are 18-in, 42.7-lb steel channels with 1-in thick by 8-in wide plates welded along the top and bottom flanges of the channel. The beams are bolted at each end to brackets welded to the steel formliner inside the reactor support pedestal. Hanger rods, about 10 ft long by 1 3/4-in diameter, are supported from the beams on stacks of disc springs which will compress about 2 in under the design load. Support bars are bolted between the bottom ends of the rods hanging from each beam. Grid plates rest on the support bars between adjacent beams. Because a single-piece grid would be too heavy to handle in the limited work space and would not allow access to individual CRD housings, each grid is designed to be assembled in place utilizing two grid plates, the clamp and the bolt. The grid clamps also provide loose lateral guidance. Beveled, loose-fitting ends on the support bars prevent large bending moments in the hanger rods if the supports are ever loaded laterally. NMP Unit 1 UFSAR Section VII VII-40 Rev. 25, October 2017 When the support bars and grids are installed, a gap of about 1 in at room temperature (approximately 70°F) is provided between the grid and the bottom contact surface of the CRD flange. During system heatup this gap is reduced by a net downward expansion of the housings with respect to the supports. In the hot operating condition the gap is approximately 1/4 in. The spring pivots at the top and the loose-fitting support bars at the bottom are designed to assure practically no resistance to horizontal movements at the lower end. It should be noted that only one grid module and four hanger rods on two adjacent structural beams will be affected in any single CRD housing failure. The American Institute of Steel Construction (AISC)* Specification for Design, Fabrication and Erection of Structural Steel for Building was used in the design of the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses were taken as 90 percent of yield, and the shear stress as 60 percent of yield. These are 1.5 times the corresponding AISC allowable stresses* of 60 percent and 40 percent of yield. This stress criterion is desirable for this application and adequate for the "once in a lifetime" loading condition. The support system provided is removable and replaceable, one module at a time, to permit access to and removal of the CRDs, position indicators, and in-core instrument components for inspection and maintenance. To gain access to any CRD housing, it is necessary to remove the grid clamp below the CRD housing of interest and remove the two grid plates. 2.1 Loads and Deflections In the event of a housing failure, the weight of the separated housing, CRD and blade, plus the force of 1250 psig pressure acting on the 6-in diameter of the housing, will produce a force of approximately 35,100 lb. To determine the design maximum dynamic deflection () of the CRD housing support system, an impact factor of 3 is substituted in the formulas:

• Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section VII VII-41 Rev. 25, October 2017 Where the maximum value of "h" is 1 in. The resulting total force becomes 105 pounds when the initial gap is conservatively assumed to be 1 in. This total force (105 lb) was then applied as a static load in standard design formulas. The calculated total deflection in the springs and supports, together with the 1-in gap, limit the control rod movement to less than 3 in. The support beams are assumed rigid relative to the hanger rods, grid plates and disc springs and the beam deflection neglected. This assumption is considered conservative since additional deflection will tend to be compensated for a reduction of impact factor. 3.0 Design Evaluation Downward travel of the CRD housing following the design basis housing failure is the sum of the initial gap and the elastic deflection of the supporting structure under dynamic loading. The support system limits the total downward movement of the control rod to approximately 3 in, in the worst case, assuming a cold clearance gap of approximately 1 in. When the reactor is hot and pressurized, the total deflection is about 2 1/4 in or less. Thus, the control rod ejection following a housing failure will always be less than one normal 6-in drive notch movement and is expected to be less than one-half notch movement. The nuclear transient from sudden withdrawal of any single control rod through a distance of one normal drive notch (6 in) or less, at any position in the core, does not result in any damage to the fuel or reactor coolant system (RCS). The CRD housing supports are in place any time the reactor is to be operated. The housing supports may be removed when the reactor is in the shutdown condition even when the reactor is pressurized, since all control rods are then inserted. In the unlikely event that a control rod is ejected under the shutdown condition, the reactor will remain subcritical since it is designed to be subcritical with any one control rod fully withdrawn at any time. NMP Unit 1 UFSAR Section VII VII-42 Rev. 25, October 2017 The CRD housing support induces no horizontal or vertical stress in the housing, since the support hanger rods are pivoted at top and bottom for free horizontal movement and the supports do not come into vertical contact with the CRD housings, except during the postulated accident conditions. 4.0 Tests and Inspections The CRD housing support will be in place any time the reactor is to be operated. Sections may be removed to permit maintenance on CRD mechanisms. Whenever maintenance or other work on the system has been performed the support structure will be inspected to assure proper installation before the reactor is returned to operation. F. FLOW RESTRICTORS 1.0 Design Bases The flow restrictors are designed to serve as constrictions in each main steam line (MSL) as close to the reactor vessel as practical, which will accomplish the following in the event of complete severance of a MSL: 1. Limit loss of coolant from the primary system to the extent that the coolant level in the vessel will not fall below the point where core cooling will be ineffective. 2. Reduce the amount of moisture carryover before closure of the main steam line isolation valves (MSIVs). 3. Reduce the possibility of forming water slugs of high velocity in steam line. The design accident under which the flow restrictors are evaluated is the postulated complete severance of a MSL outside of the primary containment. Because of rapid depressurization a steam-water mixture will leave the vessel. 2.0 System Design The flow restrictor is a simple Venturi tube welded into each MSL between the reactor vessel and first isolation valve. The restrictor has no moving parts and is located as close to the reactor vessel as practical. The ratio of the Venturi throat area to steam line flow area is approximately 0.33, which NMP Unit 1 UFSAR Section VII VII-43 Rev. 25, October 2017 results in an irreversible pressure drop of 8.6 psi at 4.0 x 106 lb/hr. The irreversible pressure drop is approximately 10 percent of the total pressure drop. This design limits steam flow in the severed line to about 200 percent of rated, yet it results in negligible increase in steam moisture content during normal operation. The flow restrictors as part of the MSLs have the same design conditions of pressure and temperature and are designed in accordance with the ASA Code, Section B31.1-1955. They are also designed to withstand the maximum pressure difference expected following complete severance of the MSL. 3.0 Design Evaluation In the event of a steam line break downstream of the restrictor, flow chokes in the decreased area by a two-phase mechanism, similar to the critical flow phenomena in gas dynamics. This limits the steam flow rate, thereby reducing the amount of coolant blowdown. Pressure surges caused by the water-steam slugs impacting the flow limiter are within the design pressure while beyond the flow limiter, the velocities are reduced and pressure surges are of no consequence. Analysis of the MSL rupture (Section XV) shows that the amount of water leaving the system before the isolation valves are closed can easily be tolerated from the standpoint of activity released to the environs. The coolant loss does not cause the core to become uncovered. Thus, the use of restrictors and the isolation valves limits the coolant loss to the point where feedwater flow maintains core coverage. Tests were conducted to determine the final design and performance characteristics of the flow restrictor, including maximum flow rate of the limiter corresponding to the accident conditions, irreversible losses under normal operating conditions, and discharge moisture level. These tests resulted in the following conclusions. 1. The nozzle (restrictor) performance is in agreement with existing ASME correlations. 2. Unrecovered P is consistently about 10 percent of the total nozzle differential and is increased a small amount by the moisture in the steam. NMP Unit 1 UFSAR Section VII VII-44 Rev. 25, October 2017 3. Operation at critical throat velocities is stable and predictable. 4.0 Tests and Inspections Since the flow restrictor has no moving parts and forms a permanent part of the MSL, no maintenance or testing is required. Very slow erosion may occur with time, but the slight enlargement should not affect the restrictive qualities. G. COMBUSTIBLE GAS CONTROL SYSTEM 1.0 Design Bases The combustible gas control system (CGCS) is designed to prevent a combustible hydrogen-oxygen concentration from accumulating in the primary containment atmosphere immediately following or during a LOCA. Basically, the CGCS performs the following functions: 1. Inerting of the primary containment within 24 hr after startup and subsequent deinerting for shutdown. 2. Supplying nitrogen makeup during normal operation. 3. Providing a controlled supply of nitrogen into the primary containment following a LOCA. 4. Providing a venting capability through the emergency ventilation system. 5. When in service provides continuous hydrogen-oxygen concentration monitoring of the primary containment atmosphere. The system is capable of reducing and maintaining the oxygen content of the primary containment atmosphere below 4.0 percent by volume to preclude the possibility of hydrogen combustion regardless of the amount of hydrogen available. The limit of hydrogen flammability in a nitrogen-oxygen mixture is shown on Figure VII-11. The CGCS consists of two functionally independent systems: containment inerting system and the containment atmosphere dilution (CAD) system. 2.0 Containment Inerting System NMP Unit 1 UFSAR Section VII VII-45 Rev. 25, October 2017 2.1 System Design The containment inerting system is designed to purge the primary containment atmosphere with pure nitrogen within 24 hr after startup and provide makeup during normal operation. The system also reintroduces air into the primary containment prior to shutdown to allow for personnel access. The containment inerting system, shown on Figure VII-12, consists of a bulk liquid nitrogen storage tank with a capacity of 11,410 gal (approximately 1,020,500 scf), a second storage tank with a capacity of 4,000 gal (approximately 385,000 scf), a vaporizing unit rated at 300,000 scfh, and all necessary piping, valves and instrumentation. Each vent and fill line to the drywell and suppression chamber from the vaporizer has two isolation valves in series. The design of equipment and piping inside the containment and up to the isolation valves is in accordance with the applicable sections of ASME Code, Section III-B-1965, and the ASA B31.1-1955 Piping Code with nuclear interpretations. Nitrogen supply for a complete purge of the primary containment to less than 4-percent oxygen is stored in liquid form in a bulk nitrogen storage tank. This tank is connected to a steam vaporizer which can deliver 300,000 scfh of gaseous nitrogen at 50°F. Steam for vaporizing the nitrogen is provided by an electric boiler. Two methods of purging the primary containment are employed; the preferred "Continuous Feed and Bleed Method" or the alternative "Batch Blow Method." Approximately 600,000 cu ft of gaseous nitrogen are required to initially inert the primary containment. Operation of the drywell cooler fans provides for maximum mixing within containment during purging. Nitrogen required for makeup during normal operation is supplied by two redundant nitrogen supplying systems (Section VII-G.3.1). Each system can supply 0-100 cfm at 50-60°F gaseous nitrogen to the drywell and torus. Makeup is initiated manually from the control room. Normally, the nitrogen is exhausted directly to the stack by a fan rated at 10,000 cfm. The nitrogen leaves the drywell at about 6,000 cfm through the upper purge and vent line. The gas leaves the suppression chamber at approximately 4,000 cfm by the same purge and vent line through which it entered. Should the nitrogen atmosphere be significantly contaminated, it can be NMP Unit 1 UFSAR Section VII VII-46 Rev. 25, October 2017 passed through the emergency ventilation system at a rate of not more than 1,600 cfm. The system is also provided with a hardened vent path which bypasses the containment vent and purge fan. This path provides emergency containment venting capability under degraded accident conditions. A second hardened vent is installed in order to vent from primary containment the steam equivalent of a decay heat rate of 1% of the rated thermal power at the design pressure of 35 psig, to be used during a Beyond Design Basis Event (BOBE) resulting in an Extended Loss of AC Power (ELAP). The design of the system is guided by NRC Order EA-13-109 and NEI 13-02. The EA-13-109 Hardened Containment Vent System (HCVS) consists of a 10" branch line off of the 20" vent and purge line from the torus, interfacing with the line upstream of IV-201-16. The 10" pipe is routed up the Reactor building and exits through the roof. Air actuated valves IV-201.13-74 and IV- 201.13-71, located at the start of the 10" vent line, are operated via a Nitrogen based motive gas system located at the Remote Operating Station (ROS). The ROS is located on Turbine Building elevation 261', between the heater bays, along column row 6. The motive gas system can be operated from either the Control Room or locally at the ROS after initial valve lineups are made. Since venting during a severe accident will result in the potential of a combustible mixture in the pipe after venting, an Argon purge system, also located at the ROS, is used to purge out the contents of the pipe and create an inert atmosphere. Similar to the motive gas system, the purge system can be operated from the Control Room or locally at the ROS after initial valve lineups are made. The HCVS is powered by a dedicated set of batteries located at the ROS that are designed to operate for 24 hours prior to needing to be recharged. The batteries power the solenoid operated valves of the motive gas and purge system as well as instrumentation including vent line temperature and radiation indicators, purge bottle pressure indicator, valve position switches, a voltage monitor, a purge a data recorder and indicating lights in the control room. The vent line is prevented from becoming a bypass leakage pathway with the use of a leak tight rupture disc set to burst above the maximum LOCA pressure, located in the line downstream NMP Unit 1 UFSAR Section VII VII-47 Rev. 25, October 2017 of IV-201.13-71. An open vent in the purge line within secondary containment also prevents a bypass leakage pathway by venting any pressure build up between IV-201.13-71 and the disc into secondary containment. In order to operate the HCVS, the purge system is used to pressurize the space between IV-201.13-71 and the rupture disc, bursting the disc and creating the vent path. 2.2 Design Evaluation By purging the primary containment with nitrogen, the oxygen content of the primary containment atmosphere is reduced to 4.0 percent or less by volume. This initial inerting of the primary containment is sufficient to prevent a flammable hydrogen-oxygen mixture from accumulating if a metal-water reaction were to occur immediately following a LOCA. 3.0 Containment Atmospheric Dilution System 3.1 System Design The CAD system is designed to limit the oxygen concentration of the primary containment atmosphere to less than 4.0 percent during a LOCA. Following a LOCA, hydrogen and oxygen may be released within the primary containment from postulated metal-water reactions and from radiolysis. The initially inerted primary containment prevents the combustion of hydrogen evolved from a metal-water reaction. However, radiolytic decomposition results in the release of both hydrogen and oxygen. The CAD system functions by adding nitrogen to the primary containment atmosphere as the radiolytic formation of oxygen occurs. Oxygen concentration is, therefore, diluted to remain below 4 percent by volume. Since the radiolysis rate decreases with time as a result of fission product decay, the required nitrogen addition rate will also decrease with time. Nitrogen required for CAD system operation is supplied by two redundant nitrogen supply systems. Each nitrogen supply system (Figure VII-12) consists of a storage tank, vaporizer, electric heater and all required piping, valves and instrumentation. Discharge from either system is to the normal containment inerting system piping downstream of the two isolation valves. The preferred nitrogen supply system used under accident conditions utilizes the same nitrogen storage tank that supplies the containment inerting system. This storage tank contains NMP Unit 1 UFSAR Section VII VII-48 Rev. 25, October 2017 11,410 gal of liquid nitrogen. The redundant nitrogen supply system utilizes a 4000-gal capacity storage tank. Each nitrogen supply system is designed to supply 0-100 cfm at 50-60°F gaseous nitrogen to the drywell and suppression chamber. The system also employs a containment venting capability. This capability provides for venting the primary containment through the emergency ventilation system or to the main condenser. It also provides vent paths with pressure control valves to ensure that the downstream pressure does not exceed 0.5 psig. The level of radioactivity in the atmosphere inside the primary containment is monitored, should the necessity arise for venting after a LOCA. Two redundant hydrogen and oxygen sampling systems (Figure VII-13) are also an integral part of the CAD system. When in service they continuously monitor the hydrogen and oxygen concentrations within the drywell and suppression chamber to minimize sampling errors. Two sampling probes check the drywell atmosphere while two sampling probes check the suppression chamber atmosphere. A continuous indication of hydrogen concentration (0-20 percent) and oxygen concentration (0-5 percent and 0-25 percent) in the primary containment atmosphere is provided in the control room. All equipment was designed to operate in the most severe environmental conditions. The environmental qualification of the system components has been reevaluated through the equipment qualification project. CAD system operation is controlled completely from the control room. Primary containment pressure (Section VIII-C.2.0) is also monitored and displayed in the control room. The CAD system is designed as a non-safety related, non-seismic Class I system, and in accordance with NRC RG 1.7-2007. 3.2 Design Evaluation The hydrogen and oxygen concentrations as a function of time after occurrence of a design basis LOCA, calculated in accordance with the guidance in RG 1.7, Revision 2, are shown on Figure VII-14. From this figure it can be seen that nitrogen injection could begin at 7.2 hr after the accident. Figure VII-15 shows the effects of nitrogen addition as a function of time if performed after a LOCA. If nitrogen gas is added to the NMP Unit 1 UFSAR Section VII VII-49 Rev. 25, October 2017 containment, which is a fixed volume, an increase in pressure, proportional to the amount of gas added, occurs. Figure VII-16 shows containment pressure as a function of time after a LOCA assuming zero containment leakage. If nitrogen is injected, the pressure limit of the suppression chamber (35 psig) is reached at approximately 100 days (delivery of nitrogen can be continued for containment pressures in excess of 40 psig). The EOPs provide the required post-accident actions for primary containment if the CAD system is placed into operation. In the above analysis, zero containment leakage was assumed for conservative purposes. However, pressures will be less than those on Figure VII-16 as a result of normal containment leakage rates. 4.0 Tests and Inspections At least once per week the oxygen concentration is determined. Once each operating cycle all CAD systems undergo a leak test using a helium tracer, followed by repairs and retest, if required. H. EMERGENCY VENTILATION SYSTEM 1.0 Design Bases The emergency ventilation system is designed to filter particulates and iodines from the reactor building atmosphere prior to exhausting to the stack during secondary containment isolation conditions. Accident analyses indicate that the most severe release of fission products to the reactor building would be for the refueling accident discussed in Section XV. However, to be conservative, a release corresponding to the TID-14844 accident is used as the design basis for this system. The filter design basis was also evaluated using AST assumptions and the results concluded the TID-14844 is conservative and bounding. Only small amounts of particulates would be released from the containment during an accident, and approximately 6.5 percent of the total iodines are available for leakage from the containment to the reactor building. The filtering system has been designed to accommodate the expected total heat load of approximately 936 W resulting from the decay of deposited radioactive material. The filter heating values are based on a LOCA using TID-14844 fission product NMP Unit 1 UFSAR Section VII VII-50 Rev. 25, October 2017 release fractions from the fuel, and leakage rates of 1.5 percent per day for the containment, and a fan discharge rate equivalent to 100 percent of the reactor building volume per 24 hr. Adequate provisions are available to cool the filters to maintain their effectiveness. Immediate operation of the emergency ventilation system can be automatically initiated upon detection of a high radiation signal at the refuel platform when required (typically for fuel handling-related activities). In addition, the reactor building exhaust duct is monitored for high radiation by two monitors. The monitors are electrically connected in a one-out-of-two taken once logic. High radiation must be detected by at least one monitor for a period of 2 to 3 sec to produce an actuation signal. This time delay reduces spurious actuations while ensuring a valid actuation occurs. An EQ program for electrical equipment has been conducted in accordance with 10CFR50.49. As a result of this program, electrical equipment in the emergency ventilation system important to safety has been qualified to operate in the environment to which it is exposed. 2.0 System Design A schematic piping and instrumentation diagram (P&ID) of the emergency ventilation system is included as part of the reactor building ventilation system diagram, Figure VI-24. The system consists of a common supply header taking suction from the normal reactor building ventilation discharge before the inlet isolation valves, an electric heater (10 kW) located on the common supply duct, a dual bank of filters for removal of particulates and halogens, a 1,000-W heater in each filter bank, a motor-driven fan in each bank, isolation valves at the supply and exhaust of each bank, and separate discharge ductwork from each fan provided with independent flow nozzles and flow control instrumentation. The duct work and other equipment are designed to operate in a reactor building environment of 150°F, based on the worst-case accident for the secondary containment. The ductwork is also designed to withstand 0.5 psig, negative or positive. This pressure is based on the pressure head developed by the fans. Each filter bank has a rated flow capacity of 1,600 cfm with the building at negative pressure of 0.25 in W.G. relative to the outside. One of the two filter banks is considered as a full-capacity spare, since each is capable of one complete NMP Unit 1 UFSAR Section VII VII-51 Rev. 25, October 2017 change of air in the reactor building per day and performing the required filtering duty. However, upon a high radiation signal from the reactor building normal ventilation system or refueling platform, both loops will be activated until one loop is shut down by an Operator. The filter banks have a common supply header, but have independent exhaust ductwork from each fan, thereby satisfying the single failure criterion. A system isolation valve is provided in the common supply header (i.e., connection of normal and emergency ventilation systems). An isolation valve is also provided in the supply duct to each filter bank. Provisions have also been made to admit turbine building air to each filter bank for cooling, should a filter become overloaded or damaged, and removal from service becomes necessary. If a filter system were shut down after an accident, the other fan could draw outside air through the idle filter. A maximum of 60 cfm of turbine building air can be admitted to the inlet of the idle filter. The fan associated with the operating filter draws the turbine building air through the idle filter by means of a valved interconnection located between the filter outlets and the fans. An electric heater is provided in the common supply duct of sufficient capacity (10 kW) to reduce the relative humidity from 100 percent to 70 percent. Since both trains actuate simultaneously, the humidity can be as high as 80 percent for a time period of up to 30 min. This will allow for Operator action to turn one train off. However, filter efficiency remains at 95 percent or greater as demonstrated by testing required by Section 3.4.4.c of the Technical Specifications. This assures that filter efficiencies remain high since, in the extremely high humidity ranges (approximately 95 percent), efficiency is adversely affected. Each filter bank includes the following, in sequence of treatment: 1. A high-efficiency particulate absolute (HEPA) filter, water resistant, capable of removing 99.97 percent minimum of particulate matter which is 0.3 micron or larger in size. Filter design is fire resistant for temperatures up to 500°F. 2. A charcoal filter, with activated and specially impregnated carbon, which was originally specified to be capable of removing 99.0 percent of radioactive methyliodide and other iodine forms. Actual iodine removal efficiency is verified by testing in accordance with the Technical Specifications. Tests(2,3) have demonstrated that impregnated charcoal filters NMP Unit 1 UFSAR Section VII VII-52 Rev. 25, October 2017 are capable of adsorbing organic iodines up to at least 3,500 µg CH3I/gm charcoal, which is much greater than the design basis load of about 936 µg CH3I/gm. The filters are cooled by the normal air flow of 1,600 cfm. Provisions have been made for additional cooling should the situation arise. A 1,000-W heater is provided for each charcoal filter to prevent condensation when the system is first placed in service. Tests have shown that high efficiencies can be maintained even under reasonably high humidity conditions (~70 percent).(4,5) The 10-kW duct heater reduces humidity within the ranges covered by these tests. 3. A second HEPA filter, following the iodine filter, is also capable of removing 99.97 percent minimum of particulates larger than 0.3 micron. This filter is provided to collect any particles which might become dislodged from the charcoal filter. The emergency ventilation system with gas cleaning equipment is placed in operation automatically when the normal reactor building ventilation system is automatically shut down and isolated. Isolation of the reactor building ventilation system and startup of the emergency ventilation system occur upon high radiation in the discharge line to the normal system exhaust fans, or from high radiation at the refueling platform during refueling operation; both loops will be activated until one loop is shut down by an Operator. The system can be manually initiated. This system can be used as an alternate discharge system for reactor vessel venting if flooding of the pressure suppression system and the reactor vessel is required. 2.1 Operator Assessment For each of the two emergency ventilation fans a control room, panel-mounted control switch with indicating lights is provided. Position indicator lights for fan inlet and outlet valves are furnished. Separate flow control instrumentation monitors the discharge flow rate in each independent exhaust duct. Low flow alarms are annunciated in the control room. Differential pressure switches across each of the absolute filters and the charcoal filters monitor plugging of these filters and provide a high differential pressure alarm to the control room annunciator NMP Unit 1 UFSAR Section VII VII-53 Rev. 25, October 2017 and the Station computer. A filter inlet high temperature alarm is also provided for each system. All sensors are in accessible locations and are provided with suitable valving for in-place testing at any time. 3.0 Design Evaluation The emergency ventilation fans discharge a volume equivalent to 100 percent of the building volume per 24 hr through high efficiency particulate and charcoal filters. The filter heating values are calculated from the amount of radioactive iodine which would be available to be deposited on the charcoal filters, based on a leakage rate of 1.5 percent per day for containment. The total iodine deposited in the filters would be approximately 936 g CH3I/gm charcoal which is well within the filter design capabilities of 3500 g CH3I/gm charcoal. A bypass arrangement utilizing turbine building air is provided to assist in cooling the filters, should one bank become overloaded and have to be removed from service. To maintain the filter temperature below 500°F, only about 10 cfm of turbine building air at 100°F would be required for the decay heat load generated by the total iodine deposited. Charcoal ignition temperature is approximately 650°F. At the full heat load the bypass arrangement described above can supply 60 cfm of cooling air. With this amount of cooling air, charcoal temperature would not exceed 200°F. The capability of the system to maintain a negative pressure of 0.25 in of water with only one fan will prevent exfiltration from the reactor building. 4.0 Tests and Inspections Particulate filters are shop tested with DOP (dioctylphthalate) for a minimum removal efficiency of 99.97 percent. Immediately prior to installation each filter is thoroughly inspected for damage, tears and pinholes by illuminating the back side with strong light. Any such damage is cause for rejection. After installation the filters are tested to demonstrate that they are undamaged and properly sealed in place. For the particulate filters the test consists of injecting DOP upstream of the filter and surveying the downstream side of the filter for leaks. The DOP is introduced in a manner to provide good mixing, and care is taken to keep the mixture from the charcoal filters. For the charcoal filters, the test consists of NMP Unit 1 UFSAR Section VII VII-54 Rev. 25, October 2017 injecting methyliodide or other halogenated hydrocarbon upstream of the filters and surveying the downstream side of the filters for the material injected. The emergency ventilation system is normally a standby system which must perform only in the event of an accident. To assure that the filters have not deteriorated or lost capacity, periodic efficiency testing is performed in accordance with Technical Specifications. Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 in of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability and pressure drop are determined at least once per operating cycle to show system performance capability. Demonstration of the automatic initiation capability and operability of filter cooling (at least once per operating cycle) assures system performance capability. Test connections installed upstream of each flow element are available to utilize portable test equipment to verify the accuracy of the flow elements. I. HIGH-PRESSURE COOLANT INJECTION 1.0 Design Bases The high-pressure coolant injection (HPCI) system is an operating mode of the feedwater system available in the event of a small reactor coolant line break which exceeds the capability of the CRD pumps (0.003 ft2). A single train of HPCI, along with one emergency cooling system, has the capability of keeping the swollen reactor coolant level above the top of active fuel (TAF) for small reactor coolant boundary breaks up to 0.063 ft2 for at least 1000 sec. The HPCI system, with one of the two emergency cooling systems and two core spray systems, will provide core cooling for the complete spectrum of break sizes up to the maximum design basis recirculation discharge line break (5.446 ft2). Its primary purpose is to: 1. Provide adequate cooling of the reactor core under abnormal and accident conditions. NMP Unit 1 UFSAR Section VII VII-55 Rev. 25, October 2017 2. Remove the heat from radioactive decay and residual heat from the reactor core at such a rate that fuel clad melting would be prevented. 3. Provide for continuity of core cooling over the complete range of postulated break sizes in the primary system process barrier. HPCI is not an engineered safeguards system and is not considered in any LOCA analyses. It is discussed in this section because of its capability to provide makeup water at reactor operating pressure. 2.0 System Design The HPCI system utilizes the two condensate storage tanks (CST), the main condenser hotwell, two condensate pumps, condensate filters, condensate demineralizers, two feedwater booster pumps, feedwater heaters, two motor-driven feedwater pumps, an integrated control system and all associated piping and valves. The system is capable of delivering 6840 gpm into the reactor vessel at reactor pressure when using two trains of feedwater pumps. However, the design analyses assume a single train of HPCI is operating. The condensate and feedwater booster pumps are capable of supplying the required 3420 gpm at approximately reactor pressures up to 293 psig*. Above 293 psig, a motor-driven feedwater pump is necessary to provide the required flow rate. The feedwater system pumps have recirculation lines with air-operated flow control valves to prevent the pumps from operating against a closed system. In the event of loss of air pressure, these valves open, recycling part of the HPCI flow to the hotwell. The shaft-driven pump min-flow valve connecting the discharge header to the suction header fails closed to prevent redirection of large amounts of feedwater away from the reactor. HPCI flow would be reduced to approximately 2600 gpm at a reactor pressure of 1030 psig and 3420 gpm at a reactor pressure of 715 psig for a loss of instrument air event. Condensate inventory is maintained at an available minimum volume of 180,000 gal. 3.0 Design Evaluation

• 293 psig provides for system pump degradation of 10 percent.

NMP Unit 1 UFSAR Section VII VII-56 Rev. 25, October 2017 During a LOCA within the drywell, high drywell pressure due to a line break will cause a reactor scram. This automatic scram will cause a turbine trip after a 5-sec delay. Feedwater flow would be available for considerable time from the shaft-driven feedwater pump. The shaft-driven feedwater pump would coast down while the electric motor-driven condensate pumps and feedwater booster pumps would continue to operate. The coastdown time to reach 3420 gpm delivery to the core is approximately 4.8 min (Figure VII-17), since both the condensate and feedwater booster pumps will continue to operate on offsite power. The curve on Figure VII-17 shows how flow from the shaft-driven feedwater pump decreases as the main turbine is coasting down following a trip. The curve is a representation of the feedwater capability of the shaft-driven pump after a turbine trip at a set of finite conditions. The margin to reach the 4.8-min coastdown time is governed by the turbine coastdown rate and the shaft-driven pump, not system resistance such as flow control valve (FCV) position. The turbine trip will signal the motor-driven feedwater pump to start. The signal will be simultaneous with the start of the shaft pump coastdown. The motor-driven feedwater pump will be up to speed and capable of supplying 3420 gpm in about 10 sec. As a backup, low reactor water level will also signal the motor-driven pump to start. The initiation signal transfers control from the normal feedwater to the HPCI instrumentation and controller which has been continuously tracking the normal feedwater control signal. To maximize the NPSH to the motor-driven feedwater pumps when operating in HPCI mode, #11 flow control valve (FCV11) for #11 motor-driven feedwater pump (FWP11) does not open if there is sufficient total feedwater flow into the reactor. FCV11 remains closed until total feedwater flow into the reactor drops below 4.5 x 106 lbm/hr (9000 gpm). This logic is bypassed if FWP12 is not running or locked out. In addition, the level setpoint setdown controller (ID66B) limits the controller output to 35 percent of maximum following HPCI actuation. Feedwater flow will continue to be provided by the shaft-driven feedwater pump during turbine coastdown. Thus, there will be a continuous supply of feedwater to the reactor. The HPCI single element control system will attempt to maintain reactor vessel water level at 65 in or 72 in (depending upon which pump, 11 or 12, respectively, is in service) with a design basis feedwater flow of 3420 gpm. A sustained high reactor water level RPS signal coincident with an open feedwater flow control valve will selectively trip the NMP Unit 1 UFSAR Section VII VII-57 Rev. 25, October 2017 associated feedwater pump. The clutch of the shaft-driven pump will also be disengaged immediately upon high reactor water level. Independent of the original high water level trip installed to meet NUREG-0737 commitments, a nonselective backup trip of the motor-driven feedwater pumps will be actuated if reactor water level remains high. Should the reactor water level reach the low level scram setpoint, the motor-driven pump that tripped on high reactor water level will restart. Necessary feedwater pump recirculation is provided to allow for continued pump operation with the FCV closed. As feedwater is pumped out of the condenser hotwell, through the selected equipment of the condensate and feedwater systems and into the reactor, the condenser hotwell level will fall. Since condensed steam from the turbine no longer replenishes the condenser hotwell, condensate will be transferred from the CSTs to the hotwell for makeup. The FWS system pumps operate on 4160 V. When the plant is in operation, the power is supplied from the main generator through the Station service transformer when the generator is on-line and connected to the grid. When the main generator is off-line, the feedwater pumps are supplied with normal offsite power from the 115 kV system through the reserve transformers. If a HPCI initiation signal should occur, all HPCI/FWS system pumps would start immediately with two feedwater pump trains available for HPCI injection using the single-element feedwater control system for reactor vessel level control. If a major power disturbance were to occur that resulted in loss of the 115-kV power supply to the Nine Mile Point 115-kV bus, power would be restored from a generator located at the Bennetts Bridge Hydro Station. This generator would have the capacity of supplying approximately 6,000 kVA which is sufficient to operate one train of HPCI/FWS system pumps. If HPCI initiation were to occur, the preferred feedwater train pumps (feedwater pump 12, feedwater booster pump 13, condensate pump 13) would start. The nonpreferred train pumps would be electrically locked out on a LOOP and not start until the Operator manually reset the lockout by placing the backup pump control switch in the trip or close position. If a preferred pump train pump control switch had been manually locked out prior to the LOOP, it would remain locked out and the nonpreferred train backup pump would automatically start on HPCI initiation. If both the preferred and backup pumps are running, the preferred pump would remain in service and the backup pump will trip. The use of a Bennetts Bridge hydro generator, while NMP Unit 1 UFSAR Section VII VII-58 Rev. 25, October 2017 not equivalent to an onsite emergency power source, provides a highly reliable alternate offsite power supply for the HPCI function of the FWS system. 4.0 Tests and Inspections Tests and inspections of the various components are described in Section XI - Steam-to-Power Conversion. J. GAS ACCUMULATION 1.0 Nine Mile Point response to GL 2008-01 The Core Spray and Containment Spray system reviews performed to address GL 2008-01 determined that these systems were not susceptible to voiding, that has not already been considered in the system design. No susceptible locations on either of the systems suction piping were identified during the Ultrasonic inspections that were performed. For both systems, a significant portion of the discharge piping is voided by design. No susceptibility was found in the keep-fill portion of the core spray system. For Shutdown Cooling, several improvements to the system filling and venting procedure were implemented as part of the GL 2008-01 review. The Shutdown Cooling (SDC) procedure ensures that a rigorous filling and venting process is utilized. Based on the system modifications, procedure changes implemented and UT inspections performed, it has been determined that the SDC procedure is adequate to ensure the system is being maintained full of water. Review of the other proposed changes per TSTF-523 have determined that these changes are not applicable to NMP1. K. REFERENCES 1. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-24, March 2017. 2. R. D. Ackley, R. E. Adams, and W. E. Browning, Jr., "Removal of Radioactive Methyl Iodide from Steam Air Systems," ORNL-4040, January 1967. 3. "Connecticut Yankee Charcoal Filter Tests," CYAP-101, December 1966. NMP Unit 1 UFSAR Section VII VII-59 Rev. 25, October 2017 4. R. D. Ackley et al, op cit. 5. Refer to (4), p VII-49. 6. General Electric Report NEDO-10527, "Rod Drop Accident Analysis for Large Boiling Water Reactors," March 1972. 7. "Supplemental Reload Licensing Reoprt for None Mile Point 1 Reload 24 Cycle 25," 002N6949, Revision 0, March 2017. 8. NRC Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," January 11, 2008. NMP Unit 1 UFSAR Section VII VII-60 Rev. 25, October 2016 TABLE VII-1 PERFORMANCE TESTS Pressure Pressure Test Insertion (Atmospheric) (1,000 psi) Guideline P Across Piston, psi 82 80 90 15 Average Velocity, psi 3.26 2.9 3 0.3 Withdrawal P Across Piston, psi 120 100 110 10 Average Velocity, psi 3.0 3.3 3 0.3 CORE SPRAY SYSTEM fOSHIJTOO!tj ro:t.INGSYS11H SEH. "" I i I REACTOR or ., I l FROM Sf>Rlll RAW Pl.M' CCNf:rSTE I I TOSHJTOOllN COCllNG WITERSEll. _!!I'S_-; El--FIGURE VII-1 UFSAR Rev1s1on 17 October 2001 THIS FIGURE HAS BEEN DELETED FIGURE VII-2 UFSAR Revision 17 October 2001 CONTAINMENT SPRAY SYSTEM TO rn"""'°' "'"'" "'" INTll<<E TUNtEL ... WATER -TOCDRE SPRillY fU4P CX>>llAlf&IENT ''"""" toNTAINoolf;NT -*-REACTO El *---... ' ' ' ' El *----. ' ' PRIMARY LODP ORY WELL "" WAT£R ..... _J 9 kl JOINTS 8"'U. JODllS 9 9 IW.LJDlNTS ' TO DISCHllAGt ,_, COMTAINM£Nl "'"""' "" "' """' INTAKE ....... TO VASTE "'"""' "'""' 1

• STRAlt.'ER BASkETS REICIYEO AN:! APIS MAOEOBSOLETEINl'IJO.PICG.N\*q&*ll!5 J. CONTAINMENT SPRAY ISOLllTU>><I YN..VE POWER AND FAILS A!HS ON LOSS OF A[ll. 2. ORIFICE PLATE REMOVED. FIGURE VII -3 UFSAR Rev1s1on October 2003 .. , WATER "'"' '"'" INfll<<E ""'" 18 Nine Mile Point Unit 1 FSAR FIGURE VII-4 THRO FIGURE VII-5 FIGURES VII-4 THRO VII-5 HAVE BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 LIQUID POISON SYSTEM DEMltEMUZED WATER FROM STORl\GE TAN< I HOSE CDNIECTOR FOR H...TERNA.TE BORON IMJECllON LT 1------EJ L ___ Ll\i11-I SYHBOl.S EV -EXPt..OSIVEVH...'IE CCIA<tHlENT CONTINJITY MUGffT "--L..o. L..C. 1" HOSE CotftECTOR FOR ALTERNATE BaD4 INJECTIDH REACTOR L..C. FIGURE VII-6 UFSAR Rev1s1on 16 November 1999

( C> ID ... Ml NIMUM ALLOWABLE SOLUTION TEMPERATURE \ \ c::> :r ... '""" ' C> N ... C> C> ... "" C> co c::> ID c:1> No1.in1os FIGURE Vll-7 C> &I') C> :r C> M C> N C> .......... ... C> C> :r . z 0 I-::> ..J 0 C/) z w I-< 0 a:i < I-z w 0. ::? ::> 0 0 C/) I-z w u. w 0. I-::t 0 w 3: UFSAR Rev. 14 (June 1996) Nine Mile Point Unit 1 FSAR FIGURE VII-8 THIS FIGURE HAS BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 ( TYPICAL CONTROL ROD VELOCITY LIMITER n 't:L SUPPORT CASllNG UJRE SIJPPllllT PLATE 22-1/16 TO ACTIVE FUEL ZONE FIGURE VH-9 UFSAR Rev. 14 (June 1996) .. P * . . :* ._ (1 * * . *. *; fl : '"" . *. "* , "" .. * ' . DISC SPRING ... CONTROL ROD DRIVE HOUSING SUPPORT Figure 11-23 SUPPORT BAR . . ' . .-.. . ' .1.: . . . . REACTOR VESSEL : t SUPPORT PEDESTAL * ':-FIGURE Vll-10 ... . .. .. UFSAR REV. 16 (NOVEMBER 1999) 11-48 1-z UJ 80 u 60 0:::: w a. -z UJ 0 0 0:::: 40 0 :::c 20 0 25 75 . ---* .... .-,,;._, HYDROGEN FLAMMABILITY LIMITS OXYGEN IN ORIGINAL ATMOSPHERE (PERCENT) 20 15 10 5 0 I I I I " I I -f I I \. "' co FLAMMABLE ol RANGE N I \ I I ) I 80 85 90 95 100 NITROGEN IN ORIGINAL ATMOSPHERE (PERCENT) ' FIGURE Vl1-11 UFSAR Rav. 14 (June 1916) COMBUSTIBLE GAS CONTROL SYSTEM JNSTilt.l'EHTli AHO CIWTROLS F£11RED IN l'lACf FIGURE VII-12 UFSAR Rev1swn 25 October 2017

z: w (j) (j) (_J z _J o_ :z: <I (j) N 0 I N I :--: i I : I I : I

-';fl. :I l c 0 ... c Q) u c 0 0 Figure VII-14. Controlled Hydrogen and Oxygen Concentrations-Inerted Containment -*--oxygen ------Hyt:lrogen I N2 Injection B.egins at 7.2 Hours . , . .. .. ; .. .... -:: .... .... .,,. .. .... ..... -. -.......... .... .......... .. .... ....... .. .... : : : : .................... .. ,. : ; : 1, .. . ,, :. : : .. , I .. "'"' ' .. , .. I Drywell I ,,, ,. ,. ,. ,. / / I Wetwell I .... * .. 0.01 0.1 10 100 Time Aftar LOCA (days)

• c;; Cl. " ..; ... 'D c C'll LL <o co 1ii 't; .!. ... c Cl). E c 1! c 0 0 .B 'D Cl) 'D 'D <:( c Cl) Q e .... z Figure VII-15. Integrated post-LOCA Nitrogen Volume Requirement-Inerted Containment 600,000 400,000 200,000 0 0 20 40 60 80 100 Time After LOCA (days)

-C> l -. e :I II) e Q. ... c Q) E c s c 0 0 Figure VII-16. Containment Pressure Following CAD Actuation -Inerted Containment --With Nitrogen Injection ******Without. Nitrogen Injection 50 40 Repressurization Limit -------------------------------------------------------------------------------------------------------L: 30 20 / 10 _______ ----I _____ _ 0 -10 .__ __ _,_ _ __.__.....__.__._...._._._.__ __ _,_ _ __.__.....__.__._...._._._..__ __ __._ _ _.__.....__.__._..._._.__. 0.1 10 100 Time After LOCA (days) 0:: :r:: '* (f) rn _J c 0 -2 :.'"" :':::'. u 0 o_ CJ u :;:: £ Li... 9 8 6 5 4 3 2 Feedwoter Delivery Cupobility of '*1.3 Rec1ctor Feed Pump during Turbine Coostdown Time ofter Turbine frip for 1000 PSIG Reoctor f'ressure, 100/ Discharge Valve Position, *:dh Recirculation \/(Jive Opening at 3.25 MLB; HR T ---! Q t 2 3 Time after Trip tvlinutes 4 5 6 VII--17 R ')' ..__,r 8'./lSlr.Jn c_'--;2Dl 1

Nine Mile Point Unit 1 UFSAR FIGURE VIII-1 FIGURE DELETED THE INFORMATION ON THIS FIGURE IS NOW FOUND ON TABLE VIII-4 SHEETS 1 THRU 3. UFSAR Revision 16 F VIII-1 November 1999 REACTOR PROTECTION SYSTEM ELEMENTARY DIAGRAM ... r:* l1HOP) MAIN STfAM LINE HIGHR.t.DtATlON ANH OPEN HI RAO. OUMP VOLIAE HI WATERLEVE!J. "'""'°'°" """'"' LO RfACTOR

• WATI;RLE\IEL HI,.,.,_ .......,., TUREl.'l'RIP !RM LIP SCALE
• 96'JI. SCALE ""' 1112119 ,,. T'""' +/-1e1-'A MAIN STEAM 195 PSID f .f:1RM112 IUPSC.tJ.E """ a..osEDIN RUN MODE ONLY """"} W UP SCH.! r !-* -* . ,..,. I,,_., CS-"1 111-tl1 IM.!N STEAM LINE 11 !SOI.. VN..\11! CLOSURE LS ClOam VAL\1! lS 8f-119'1. OPEN CLOSl!!D IN 1:iR!iWI.& RUN MODI!. ONLY i"<M8 ![IRJM.13 wuP SCALE 11IG48 l(RHtBC) 11K39 UPSCAJ..E """'"1 tll<23 I R.me r-£J APRM UP SCALE* 12" OF RATm 0 1tle% REACTOR MCIRC FLOW REACTOR HIGH PRESSURE OPEN Rp !11-PSIG f REACTOR LOW' LON WATl!R LML OPEN Rt REACTOR LOW LOW LOW WATERLEVa. RE18 OPEN HIGH RL\CTOR WATER I.FIB.OPEN Rl!BS" i ! 111KSS l:= F """' I E HlC3H lEVEl SCRAM DISC. VOL RCIRC OPl!N YOL !E <15 GAL f REACTOR PRESSURE! ?'&se PSIG + & C CONTA.CTS &-11 OPEN L,. I L: HIC2!8 11Kte 19<26A """" *a.ow 1. ALL. VALUES NU! Tt!CH. LIMITS 2. All. LEVELS AR! TO lf'C>ICATOR SCALe.
• MUST BE DE-ENERGIZED FOR A SCRAM **""RPS SCRAM LOGIC RB.AV SOP-3.l(N PANEi..} FIGURE Vlll-2 UFSAR REVISION 20 OCTOBER 2007 PROTECTIVE SYSTEM TYPICAL SENSOR ARRANGEMENT ,-------------------------1 I I I I I REACTOR PROTETION I I """" I I I I I I I I TURBJIE1'RIP I I I CIRCUIT I I I 1 ,-----1 I I i I I I I : I t ' l I : : 1 I < l I I : l '-I-I----, I I ,--I I I I I I I L__ I ' ' ' ' ' : ! : ! : i POWEROl"EAATED PRESSUREF\B..D:FVALVES I I i ! i I I I I I I I I I & 0 I I I I ;u I o I 11 ;;! 11 l' 11 11 11 11 11 11 COMPENSo\.TEDVESSEl. lEVEl...SYSTEM 11 CONDENS1HG ---+-----------------------------1 I I I I I I I I I I I I I I I I I I

I I I I I I I I I I I I I I I I I I I I I I L-------------------------1 I COMPENSATED VESSEL LEVELSYSTEM12 I I I I I I I I I I I I I COlUMN #12 NOT SHOWN (IDENTICAL TO #1 I) FIGURE VJil-3 UFSAR REVISION 17 OCTOBER 2.001 REACTUR :FSAR FIGURE V.lll-4 RECIRCULATION FLOW AND TURBINE CONTROL SET POINT CONTROL sYNCH. (MANUAL OR AUTO. DISPATCH) MOTOR I TO OTHER .._:----w CONTROl.LERS SPEED CONTROLLER SCOOP TUBE POSITIONER PRESSURE SPEED GOV. MECH. INITIAL PRESSURE REGULATOR M -INDUC110N MOTOR G -AC GENERATOR GOV. REG. FREQJ SPEED c C -ADJUSTABLE SPEED COUPLING TYPICAL FOR EACH LOOP F I GURE V I I I -4 UFSAR REVISION 23 OCTOBER 2013 NEUTRON MONITORING INSTRUMENT RANGES SRM IRM APRM LPRM OPERATION ,. I . ,. ... . '"' Oi! ""' T ... C) Q.. + Cl z: z: ... i== < w Q ::c t; < w ti:: . ""

• la ... :::> .... ti:: "' = ""' .... a: liW ..J i I 100 40 10 4 1 .1 .01 .otn .0001 urs a: la.I ... 0 0.. .... z: ""' (.) a: u.I Q.. SOURCE FIGURE Vlll-5 UFSAR REVISION 14 JUNE 1<1<16 I I I I I I I I I I I CONnlOL BOARD REAR _______ L ______________ _ I ( ) REFUa. INSTRUMENTATION SCRAM PROHllllT-f'All,W"E CIRCUIT 1.$ DllUNG OPERATION CA113ES SCRMI WMEN OE-*NERGUED PROTECTIVE SYSTEM SCRAM LOGIC [_ ______ _ ... IMSUSER\FSAR FIGURE Vlll-6.dgn 8/3/2007 9:27:22*AM SOURCE RANGE MONITOR (SRM) ------------------------1 ------------------------....1....--I ------------------------....1.....--,.,.,00 '" '""'" PERIOD UPSC>J...ETRIP COUNT RATE LES$ THAN 111 *counrs PER SECOND I '. ---------'-----I ' ' POSITION I I L------------------------------1------t-----------SRM MONITOR BLOCK DIAGRAM : I : I ------1--SRM-2 BYPASSED SRM--2 : I L__J : I I I ---------------------t-----_ J_ --SftM..3 : I I _______ J __ IM'l\SSE:o I I I x CONTROL ROD BLOCK LOGIC SfM-1 COUNT IV.T'l!GREATl!R 'T)W( 1111 COUNra "" SECOHO FIGURE VIII-6 UFSAR REVISION 20 OCTOBER 2007 SRM DETECTOR LOCATION II Detectors FIGURE VHl-7 UFSAR Rev. 14 (June 19H)

INTERMEDIATE RANGE MONITOR CJRMJ "'"""-"""" -1111-1 T I -----------------------------== = ==== ___ J_ _ ----------------------+ I I I I I CETECTIJI POS!TllJI )UT svnrn -----. I __L_ ' I I I !-.-I : L--------L ___ ------1-______________ ,--L----, I I I ---1------1-----' : t r------------------------lNlowtsT I I I RM -llTP'550l I t" -----11111-2 *,_1PASS ____ .. -2 0------ltHBTl'OSS DM-3 I Rl-48TPflSS IRIH r ----lllt-5 BTNSS,_ ____ ....... ._ ____ ..,..BTl'OSS I ._ ----l!IH BTrASS I o-----1 t-----IRM-SBYPASS ----------------------------+--:-:---------------------1------,--... --1 r------------------lA 18 2A 28 lRH IOllTDR IRH1Nrl>2 IM-lfCl4 IM-51H16 DIHNl>I -IOITIR PRCITECTNE SYSTEM SOWi UK: ... IE47702\MSUSER1fsviii-8.dgn 8/9/2007 9:18:27 AM ' I I I 1 I I I I I PROTEt1M STSIEll ""'!Ell IRH BYPASS 5Rfr1Cfl SVITCH I t---JJtt-38YPtlSS ------------------t L --IllHB'll'A5S,_ ___ I : r--VIC.J581Pe -?---DM emm:======== ..... : *--lflHBYPtiSS RH I ---lllf-8 B\'PtlSS Ht-8 SEE Fm.RE YDl-12 APRH IUDDllllGRAM COOlll. RID 11.lXk UliJC FIGURE VII!-8 UFSAR REVISION 2111 OCTOBER 201117 IRM CORE LOCATION

• Trips Assigned To Logic Channel 1 II Trips Assigned To Logic Channel 2 FIGURE VHl-9 UFSAR Rev. 14 (June 1996)

( LPRM LOCATION WITHIN CORE LATTICE +.++.++. +++++++ 0 0 0 0 +.++.++.+ +.+ +. +++++++++++ 0 0 0 0 0 0 +.+ +.+ +.+ +.++.+ +.+ + +++++++++++++ o o D o o o +.+ +.+ +.++.+3+.+3+.+3+ +++++++++++++ 0 0 0 2 1 2 1 2 1 + + + + + + + + +.+ +.+ + *+ +*+ +*+ +*+3+ +3++3 0 0 2 1 2 1 2 +.++.++.+3+.+3+ +++++++ +0++2+'+0

• In-Core Monitor String Location Arrangement By Quadrant 1. Rotated NW Quadrant 2. Rotated NE Quadrant ' 3. Rotated SW Quadrant FIGURE VUl-10 UFSAR Rev. 14 (June 1996)

( LPRM AND APRM CORE LOCATION Control Rod Blades 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 Chamber Thimble LPRM Chamber 00000000 00000000 00000000 00000000 00000000 00000000 00000000 00000000 TIP Calibration Tube FIGURE vm-11 UFSAR Rev. 14 (June 1996) Wl!C CIMIEL LOCAL POWER RANGE MONITOR <LPRMJ ANO AVERAGE POWER RANGE MONITORS <APRMJ I I I r-1 I I r-1 I I I I """ IQllTOR 111oi1c1msmi111NG TO APRM 4 TO APRM J ro"" : I ,......____._____..___.____.__ -"'rl----'I I rAPRM BYPASS APRJllJ 5 ;: :1.NIRM _________ _,__ __ ,,.-7 I t-------1 : I t,:#RH 8 B'l?ASS APRM 8 :1: lP SCIU INJIERATIVE 1 I : I L-I I I TR P TR P TRIP IMJPERATIYE TRIP I I LCGIC CHANIEl 2 ..... =.=-=-== ...... ".,..,_=-!_ --T _:= = = = = = = = = = =-=-= = = = = = = = : = = = = = E I I I I I I I __________ L ___ _ I -Ti ----* -----r-i--POOTECTIYE SYSTEM UlilCOWH.l IA 50W'l PIOllBIT

• FAILSAFE CIRCUIT ISEt.EllCilZEDMlllilPERATlllt CAUSES SCRAM WIEN DE *Ett:llllZED " 1111-SH& " I I I -T .. --1 lllllCATJl(j Ii.:::::==,,_ __ ,_ ... __, I i:: I '"" -----------*----1 I I I -----------i--------* I I I I I: : ____________________________________________ --------' I: I I* I I I I I I I I ' I t -APRM 2 BYPASS --.... -----1 -APRM 4 BYPASS : I .----------'I rT;-'Ji PROTECTIVE SYSTEM CIWfrlll 1 :: r--APRHI I II I BYPASS II I 1--------1 1--------------------------------------------t-1* "'f-APAM 2 BYPASS I I *-APRM 3 BYPASS I I I I -APRM " BYPASS : r-------------------------11 *== """ ...... ...... -* """ SEEFIG.Vtll-8 IRM BLOCK DIACJIAM CIJIT!Q.ROCI l!LOCKLOOIC _______ _J : I -7BTPASS : I '------------------------__ _J I ,.,,,,v:.... I PERMISSIVE I I I I I I I IFERAT!JlSOONSll.E -""55 SELECTOR SWITCH PROTECTIVE SYSTEM CIWIPEL2 APRM BYPASS SELECT(ll: SWITCH I I FIGURE Vlll-12 UFSAR REVISION 20 OCTOBER 2007 APRM 11 12 13 14 ' ' . APRM 15 16 17 18 ' . ' + + + r d *
• CID CID CID CID CID = CID CID I -*c t t t -* = = CID CID -CID CID CID CID Amplifiers I --Averaging -hannel 11 Unit -bl --+ + + r -Logic C CID CID = CID CID CID = = Amplifiers -I Averaging -Logic C_!1a -Unit --nnel 12 -* t t t -*a -= = CID CID -Chamber CID CID CID = Thimble (Typical of Four) I -----

130 120 110 100 90 0::: w 80 a. 70 Q w I-60 ct 0::: l.L 0 50 0 40 30 20 10 0 TRIP LOGIC FOR APRM SCRAM & ROD BLOCK -----= ::::=--------::::------v-----:::: :-----L.--::: c:------/ / For FRTP < CMFLPD :::.---i---/ S0 = (FRTP/CMFLPD) x S / v/ / SRen = (FRTP/CMFLPD) x SRa TECH. SPEC. SCRAM / // v Where: (Analytical Limit) Sn & SRen = New Scram and Rod Block Lines -I I lECH. SPEC. ROD BLOCK / (Analytical Limit) S & SRa = Scram and Rod Block Lines shown above I FRTP = Fraction of Rated Thermal Power I CMFLPD = Core Maximum Fraction of Limiting Power Density I I I I I I lECH. SPEC. SCRAM (Analytica IUmil)(S) lECH. SPEC. ROD BLOCK (Ana lytical Limit) (s.eJ I I ROD BLOCK SETPO INT FLOW UNIT UP SCALE ytical Limit) ROD BLOCK (Anal / I" v 0 10 20 30 40 50 60 70 80 90 100 110 120 % RECIRCULATION FLOW FIGURE Vlll-14 UFSAR Revision 17 October 2001 ...a co = -" PROTECTOI/? T'Ul!JE £XP.t.05tVE YALVE l--f!HOl--1:4-. /JAIY£ IN0£XING MECHANISM ,H!t"MoWA)W l[_-m-vovr.e /N-clXiV e..ea.LE -p CONAJICTOAS 1. PIJRING R£H'l'Ai AN.0 /HMRTION .SECTION A *A OVRIN6 OPERATION ro .SU:T/ON IJ * .8 F.LUX WV/7 AGTl!S>-1'TM'6r AW CM!IMlllR.3 ..qf"A ASSEMS£Y NAY AND AlllWMllMNWNr 71:> .. &WTl*NNRO ,_JllOM .$/W6 MC> I / -t ,, > < m ,, (/) -z (') z I n 0 ,, m "ti ,, 0 °' m ROD PATTERN DURING STARTUP II Rods Withdrawn Before Criticality Is Achieved ROds Withdrawn To Bring Reactor Critical FIGURE VUl-16 UfSAft Rev. 14 (June 1991) -I Cl) QO co Cl) 0 u -a::: UJ 3:: 0 a.. a::: 0 I-u < UJ a::: . ./ RADIAL POWER DISTRIBUTION FOR CONTROL ROD PATTERN SHOWN IN FIGURE VI 11-16 10.0 5.0 1.0 0.5 0.1 0.05 PERIPHERAL ROD I 1\ : \ I \ : \ CHAMBER \ LOCATION \ \ \ 0.01...__ ___ ......... _____ __,_ __ __..i,...._ ________ ....... ___ ___, 0 5 10 15 20 25 30 DISTANCE FROM CORE BOUNDARY (fuel assemblies) RQURE Vltl-17 UFSAR Rev. 14 (June 199') DISTANCE FROM WORST CONTROL ROD TO NEAREST ACTIVE IRM MONITOR . (f) e Channel Assumed To Be Bypassed Worst Location For Control Rod Withdrawal With Respect to IRM Response Capabilities o1 =Distance From Worst CQntrol Rod To Nearest Active IRM Channel In One Bus=111.8 Assemblies = Same As Above=11. II Fuel Assemblies FIGURE VIII-18 UFSAR Rev1s1on 16 November 1 qqq MEASURED RESPONSE TIME OF INTERMEDIATE RANGE SAFETY INSTRUMENTATION 100 10 0.1 .01 .01 0.1 1.0 10 P.ERIOD -Seconds FIGURE VUl-19 UFSAR Rev. 14 (June 19H) , 16 12 10 8 6 'fft. c II 0 -;:; "' > 41 c 2 :E a::: c.. < 0 II 8 -10 ENVELOPE OF MAXIMUM APRM DEVIATION BY FLOW CONTROL REDUCTION IN POWER , All From CH. 2 I I / / /' CH. De ts CH. De ts /' 2 A&C 1 A&C 6 B&C 5 B&C 3 A&C II A&C v 7 B&C 8 B&C I Percent Rated I 80 60 110 I Flow "' All From CH. 8 ""' I FIGURE VUl*20 UFSAR Rev. 14 (June 1991) ENVELOPE OF MAXIMUM APRM DEVIATION FOR APRM TRACKING WITH MAXIMUM CONTROL ROD WITHDRAWAL .,,. c .e 16 12 10 8 6

• 4 > GI Q =: a:: 2 A. < 0 4 8 / CH. 8 v I I I " . CH. 4 J 4 I ,, I ""L ./ ----. I I CH.2 A&C CH. 6 B&D CH.3 A&C CH. 7 B&D Rod Position I I ' 8 I Feet Withdrawn 2 2 CH.1 A&C CH.S B&D CH. 4 A&C CH. 8 B&D 10 6 12 CH.& FIGURE VIII-21 UFSAR Rev1s1on 16 November-1 qqq

i MAIN STEAM LINE RADIATION MONITOR ! > I I I I t I I I I 11 _____ J ---------1 I I >---+----! : FIGURE VIII-22 UFSAR REVISION 17 OCTOBER 2001 REACTOR BUILDING VENTILATION RADIATION MONITOR FROM REACTOR BUILDING REACTOR BUILDING VENT LINE SEE FIGURE VI-24 REACTOR BUILDING ----CONTROL ROOM DOWN SCALE TRIP ALAAM IN DI CA TOR ANO TRIP UNIT UP SCALE TRIP DATA LOGGER I MONITOR 1 ------NOT UP SCALE MCl-llTOR 2 NOT LF SCALE LOGIC BLOCK DIAGRAM EMERGENCY VENTILATION PROHIBIT TO STACK REACTOR BUILDING CONTRQ... ROOM IND I CA TOR AND TRIP UNIT DATA LOGGER UP SCALE TRIP I I I I I I I I I I I I I I I I I I I -----lNirIA TE TI£ FOLLOWING1 CLO SE NORMAL VENTILATION ISOLATION VALVES NORMAL VENTD...ATION FANS TRIP STAR T EMERGENCY VENHLATICW FANS DOWN SCALE TRIP ALARM AOURE Vnt-23 UFSAR Rev. 14 IJune 19911 FROM STEAM JET AIR EJECTOR OFF GAS LINE SEE FIGURE XI-3 OFF GAS SYSTEM RADIATION MONITOR ELECTRIC Pl TO MA[N CONDENSER FROM STACK EFFLUENT MONITOR SEE FIGURE VIII -26 2 PEN RECORDER DATA LOGGER LOGARITHMIC AMPLIFIER DOWN SCALE TRIP UP SCALE TRIP ' ---------------------------------' +--------------------------' TO LOGIC CHANNEL 2 1--------------------------------------------------------; MONITOR l NOT UP SCALE MONITOR 1 NOT UP SCALE MONITOR I NOT DOWN SCALE TO LOGIC CHANNEL 2 LOGJC CHANNEL 2 SAME AS LOGIC 1 LOGIC CHANNEL l OFF GAS LINE ISOLATION PROH!BIT LOGIC CHANNEL 2 OFF GAS LINE ISOLATION PROHIBIT OFF GAS LINE ISOLAT!ON PROHJBIT DETECTOR POWER SUPPLY OFF GAS LINE ISOLATION PROHIBIT LOGIC CHANNEL 1 DETECTOR POWER SUPPL 'f LOGARITHMIC AMPLIFIER MONITOR 2 NOT DOWN SCALE MONITOR 2 NOT UP SCALE MONITOR 2 NOT UP SCALE DOWN SCALE TRIP *-------------------------------+ i TO LOGIC CHANNEL 2 UP SCALE TRIP ' ' ' --------------------------------------------------------* ' ' --------------------------------------------------------1 i DATA LOGGER FROM STACK EFFLUENT MONITOR SEE FIGURE VIII-26 2 PEN RECORDER TO LOGIC CHANNEL 2 LOGIC BLOCK DIAGRAM LOGIC CHANNEL l FIGURE VIII-24 UFSAR REVISION 17 OCTOBER 2001

(---'\ EMERGENCY CONDENSER VENT RADIATION MONITOR s;; !§ .. =: !I gi ge *= !! H ! s I ! F10URE vtll-25 UFSAR Rev. 14 (June 1996) HIGHHJGH RADJAHON ALARM STACK EFFLUENT AND LIQUID EFFLUENT RADIATION MONITORS DR Arn OFFGAS EFFLL£NT STACK MONITORrnG SYSTEM !IXESMS l RAGEHS SYSTEM SEE FIG.VllI-26a PUMP RAGEMS SYSTEM SEEFIG.YIH-261!1 UP SCALE TAI? PARTICULATE HALIXiEN FILTER FILTER DOWN SCALE TR[P LOG COUNT RATE METER INOPERAHVE TRIP DATA Lcx:;GER ATMOSPHERIC AIRPURrE PRE i\MP AND DETECTOR POWER SlPPLY FILTER FROM AIR EJECTCR DR PRE AMP ANO DETECTOR POWER SUPPLY SEE FIGURE Vlll-24 HIGH RADIAHON ALARM OOWN SCALE AND INOPERATIVE ALARM RECORDER LOG COUNT RATE METER DATA LOGGER RECORDER DOWN SCALE TRJP OOWN SCALE ANO INCf'ERATlVE ALARM LI' SCALE TAJP H[GH RADJAT!ON ALARM UP SCALE TR[P H!GHHIGH RAIJIATION ALARM LWUIO EFFLUENT MONITOR TYPICAL FOR RADWASTE SYSTEM ANO REACTOR BUILDING COO...ING WATER SYSTEM DETECTOO LOG CU ECTOR POWER SUPPLY CllJNT RATE METER TO DATA LOGGER IN(JlERATIVf TRIP DOWN SCALE AND lltlOPERATIVE ALARM 00\IN SCALE TRIP UP SCALE TRIP HIGH RAOIATJON ALARM "' SCALE TRtP HlGH H[GH RAD!AHON ALARM FIG. VIII-26 UFSAR Rev1s10n 17 October 2001 ...Jlllll!!IL I I STACK EFFLUENT AND LIQUID EFFLUENT RADIATION MONITORS _, --r -VENT l PU1iE I I I I I JI SIM'L.E OILUJl(Jrl ..... STACK GAS SAMPLE *-"' .,,., .. I 0 . i . .. REMOVABLE SECTION FOR ORAB s-._E I INSULATED> I STACK GAS "' SAMPLE PUMP .. STACK EFFLUENT AND LIQUID EFFLUENT RADIATION MONITORS Notes OGESHS provtdes normol and occident gOHous effl1Mnt mon1too-1ng. RAGEHS pt"ovides on oux1hory sonphng locot1on. n fir FIGURE Vlll-26a UFSAR Revision 20 October 2007 SERVICE WATER DISCHARGE LIQUID EFFLUENT RADIATION MONITOR I I I I I I I I I om SIUl:E I 1 __ 1m1J!L __ I I I __ _!EIECTlll __ __J_J l _________________ I I I I I _I UNIT 1 FIG. VIII-268 UFSAR Rev1s1on 25 October-2017 TO DISCHARGE DOWN SCALE TRIP DOWN SCALE ALARM HE CONTAINMENT SPRAY HEAT EXCHANGER RAW WATER EFFLUENT RADIATION MONITOR HEAT EXCHANGER DETECTOR CONVERTER FROM RAW WATER PUMPS FDA DETAILS SEE FIG. Vll-3 HEAT HE 80-14 DETECTOR CONVERTER TO OISOiAAGE TUNNEL INDICATOR AND TRIP UNIT INDICATOR ANO TRIP UNIT '-" DOWN UP SCALE SCALE SCALE TRIP Tll!P TRIP HI RADIATION DATA DOWN HI RADIATION DATA ALARM LOGGER SCALE ALARM LOGGER ALARM. TO DISCHARGE TUNNEL HE 80-33 HEAT EXCHAN:JER CONVERTER INDICATOR ANO TRIP UNIT lJ!' SCALE TRIP DATA HI RADIATION LOGGER ALARM DOWN SCALE TRIP DOWN SCALE ALARM FROM RAW WATER PUMPS FOR DETAILS SEE FIG. Vll-3 DATA LOGGER HEAT EXCHANGER HE 80-34 CONVERTER INDICATOR AND TRIP UNIT '-" SCALE TRlP Hl RADIATION ALARM FIGURE Vlll-27 TO DISCHARGE TUNNEL DOWN SCALE IBIP DOWN SCALE ALARM UfSAR Rev. 14 (June 19") CONTAINMENT ATMOSPHERIC MONITORING SYSTEM REACTOR 1-------------------...,....----------&--tf'i __ _ T I I 8 I : 1:----;---, l__:

• ____ 1 ! ; ""' OJ)]] -=--ixi--, I: i i I f =* """ n::::::;::::r-,--v S,..Ml'L.ERElllllN EL. FIGURE Vlll -28 UFSAR REV!SlON 18 OCTOBER 2003 FSAR FIGURE Vlll-29 ROD WORTH MINIMIZER Program and sequence inputs Control rod position data System control inputs RPIS Keylock switch 1/0 Typer Process computer InpuUoutput control Output buffer Output status controls Display .,.__ __ control panel Operator controls Control rod permissives/blocks FIGURE VIll-29 UFSAR Revision 17 October 2.0.01

NMP Unit 1 UFSAR LIST OF EFFECTIVE FIGURES SECTION IX Figure Revision Number Number Section IX EF IX-1 Rev. 25, October 2017 IX-1 24 IX-2 25 IX-3 14 IX-4 14 IX-5 14 IX-6 22 IX-6a 24 IX-6b 23 IX-7 14 IX-8 17 NMP Unit 1 UFSAR Section IX IX-i Rev. 25, October 2017 TABLE OF CONTENTS Section Title SECTION IX ELECTRICAL SYSTEMS A. DESIGN BASES B. ELECTRICAL SYSTEM DESIGN 1.0 Network Interconnections 1.1 345-kV System 1.2 115-kV System 2.0 Station Distribution System 2.1 Two 24-V Dc Systems 2.2 Two 120-V, 60-Hz, Single-Phase, Uninterruptible Power Supply Systems 2.3 Two 120-V, 57-60 Hz, One-Phase, Reactor Trip Power Supplies 2.4 One 120/208-V, 60-Hz, Instrument and Control Transformer 2.5 One 120/240-V, 60-Hz, Single-Phase, Computer Power Supply 3.0 Cables and Cable Trays 3.1 Cable Separation 3.2 Cable Penetrations 3.3 Protection in Hazardous Areas 3.4 Types of Cables 3.4.1 Power Cable 3.4.2 Control Cable 3.4.3 Special Cable 3.5 Design and Spacing of Cable Trays 3.5.1 Tray Design Specifications 3.5.2 Tray Spacing 4.0 Emergency Power 4.1 Diesel Generator System 4.2 Station Batteries 4.3 Q-Related Battery System 5.0 Tests and Inspections 5.1 Diesel Generator 5.2 Station Batteries 5.3 Q-Related Battery 6.0 Conformance With 10CFR50.63 - Station Blackout Rule 6.1 Station Blackout Duration 6.2 Station Blackout Coping Capability 6.3 Procedures and Training NMP Unit 1 UFSAR Section IX IX-ii Rev. 25, October 2017 6.4 Quality Assurance 6.5 Emergency Diesel Generator Reliability Program 6.6 References NMP Unit 1 UFSAR Section IX IX-ii Rev. 25, October 2017 LIST OF TABLES Table Number Title IX-1 MAGNITUDE AND DUTY CYCLE OF MAJOR STATION BATTERY LOADS NMP Unit 1 UFSAR Section IX IX-iii Rev. 25, October 2017 LIST OF FIGURES Figure Number Title IX-1 A.C. STATION POWER DISTRIBUTION IX-2 CONTROL AND INSTRUMENT POWER IX-3 TRAYS BELOW ELEVATION 261 IX-4 TRAYS BELOW ELEVATION 277 IX-5 TRAYS BELOW ELEVATION 300 IX-6 DELETED IX-6a DIESEL GENERATOR #102 LOADING FOLLOWING LOSS-OF-COOLANT ACCIDENT IX-6b DIESEL GENERATOR #103 LOADING FOLLOWING LOSS-OF-COOLANT ACCIDENT IX-7 DIESEL GENERATOR LOADING FOR ORDERLY SHUTDOWN IX-8 DIESEL GENERATOR FUEL OIL SYSTEM NMP Unit 1 UFSAR Section IX IX-1 Rev. 25, October 2017 SECTION IX ELECTRICAL SYSTEMS A. DESIGN BASES The Station electrical system is designed to provide adequate normal and emergency sources of electrical power for normal operation and for the prompt shutdown and continued maintenance of the Station in a safe condition under all credible circumstances. To guard against the remote possibility of the loss of all electrical power from sources outside the Station coincident with an accident within the Station, two completely independent emergency diesel generator systems are provided, each having a capacity adequate to provide power to all of the loads that are deemed essential on an emergency basis. This complete redundancy in the emergency generator systems parallels that in the core and containment spray systems to assure the highest possible degree of reliability in these safety systems. The Station design provides two completely independent control battery systems for redundancy and selectivity of control power sources, which greatly enhances the reliability of essential control and protective system circuitry. Loads essential to Station safety are split and diversified between auxiliary power buses, with means provided for rapid location and isolation of faults. B. ELECTRICAL SYSTEM DESIGN 1.0 Network Interconnections 1.1 345-kV System The output of the Nine Mile Point Nuclear Station - Unit 1 (Unit 1) is transmitted over two 345-kV transmission lines (#9 line to Scriba Station, approximately 0.41 mi, #8 line to Clay Station, approximately 26 mi) where it is fed into the Niagara Mohawk Power Corporation (NMPC) cross-state bulk power transmission system. The two transmission lines occupy a common right-of-way but are physically separated and supported on completely independent NMP Unit 1 UFSAR Section IX IX-2 Rev. 25, October 2017 structures to minimize the possibility of a double-circuit outage. The lines are designed to meet or exceed the requirements of the National Electric Safety Code for heavy loading districts, Grade B. The design provides theoretical lightning performance of less than 1.17 outages per 100 mi per year. Transmission lines #8 and #9 are protected by 550-kV, 3000-amp, three-phase, 50-kA, SF6 gas circuit breakers, as shown on Figure IX-1. Each line has a capacity in excess of the full expected output of the turbine generator. Normal operation is with both breakers closed and all lines energized. Redundant protective relay schemes (per New York Power Pool (NYPP) requirements for protection of bulk power systems), including backup functions coupled with automatic reclosing of line breakers, provides for a high degree of reliability in line operation. In the event that one of the two lines is temporarily out of service (OOS), the other line is capable of carrying the full Station output at no risk to system stability. Loss of all 345-kV lines will result in load rejection of the Station's net generation output being carried at the time. If the load rejected is within the range where opening of the bypass valves will permit continued operation of the reactor, the turbine generator will continue to run and carry the Station auxiliaries. On the other hand, if the load rejected is greater, reactor trip will result and the Station auxiliaries are automatically transferred to the 115-kV reserve source. When the main turbine generator is out of service and Station power is being supplied by transformers T101N and T101S, a 345-kV backfeed can be established. Backfeed is accomplished by energizing main transformer T1 or T2 by way of 345-kV lines #8 or #9, after disconnecting the main generator links and closing in on the 345-kV breakers R915 or R925, and after taking the appropriate precautions. This configuration will step down the system voltage from 345 kV to 24 kV, and then through the Station service transformer #10 to 4160 V to energize power boards (PB) 11 and 12. The use of backfeed through the T1 or T2 transformers provides an additional source of offsite power in addition to the two 115-kV reserve feeds. 1.2 115-kV System Power for Station startup, the reserve supply to the auxiliaries, and the normal supply to selected auxiliaries is obtained from the 115-kV bus. This bus is fed by two 115-kV NMP Unit 1 UFSAR Section IX IX-3 Rev. 25, October 2017 transmission lines from remote generating stations. One line is from the South Oswego Steam Station (Line #1), approximately 12 mi away. The other line is from the Lighthouse Hill Station, approximately 26 mi away, through the J. A. FitzPatrick switchyard (Line #4). Both stations have other tie line connections into the Company statewide transmission system. Lighthouse Hill includes hydroelectric generators which have the capability of startup without power input from outside sources (Black Start). The lines are designed to meet or exceed the requirements of the National Electric Safety Code for heavy loading districts, Grade B. Each line is protected by a 115-kV, 1200-amp, three-phase, 5,000-MVa oil circuit breaker. Two redundant sets of protective relays are provided on each line for automatic tripping of the circuit breakers under fault conditions. Recognizing that most line faults are transient in nature, automatic reclosing equipment and circuitry is provided to reenergize the lines after the extremely short interruption required to clear a temporary fault. The following failure mode and effects analysis (FMEA) includes the effects of all failure modes of the 115-kV reserve bus and 4.16-kV power boards. The unit is operating at or near rated load, and all auxiliaries are being supplied from their normal sources. System or Equipment Malfunction Effect 115-kV Both lines lost 115-kV bus will be transmission due to an accident de-energized. PB 11 and lines common to the 12 will not be affected. transmission lines. PB 101, 102 and 103 will be de-energized. PB 102 and 103 will be reenergized automatically via diesel generators 102 and 103. PB 101 will remain de-energized. 115-kV Major system power 115-kV bus will be transmission disturbance resulting de-energized. PB 11 and lines in temporary loss of 12 will not be affected. input power to the PB 101, 102 and 103 NMP Unit 1 UFSAR Section IX IX-4 Rev. 25, October 2017 Nine Mile Point will be de-energized, 115-kV reserve bus. but PB 102 and 103 will Lines intact and unit be reenergized by the continues to run at diesel generators. or near rated load. Undervoltage relays will trip the line breakers (R40, R10) in 30 sec. Protective relay schemes at Lighthouse Hill will automatically clear all necessary buses at this Station and actuate an alarm at the EMS Central Regional Control Center located at Henry Clay Boulevard in Liverpool. The protective relay scheme will also initiate an automated control scheme to switch one of two generators at Bennetts Bridge to the line supplying the Lighthouse Hill Station. With the line energized to Lighthouse Hill, the Lighthouse Hill line to the James A. FitzPatrick Nuclear Power Plant bus and to Nine Mile Point Nuclear Station will be energized. The line breaker at Nine Mile Point will automatically close with a live line and dead bus, energizing one-half of the 115-kV reserve bus. This will allow the Operator to switch 102 or 103 back to the offsite generation at his discretion. The other Nine Mile Point line breaker will close only if NMP Unit 1 UFSAR Section IX IX-5 Rev. 25, October 2017 the Oswego Steam Station reenergizes the line to Nine Mile Point. Under system emergency conditions this hydro generation is reserved for the Nine Mile Point power requirements exclusively. 115-kV Low or high The offsite power transmission voltage and/or system is supervised by system frequency. the following protective relays: a. overvoltage b. undervoltage c. overfrequency d. underfrequency These protective relays are set to transfer PB 102 and 103 to their respective onsite diesel power systems if the voltage and/or frequency of the offsite system deviates beyond set limits required for the satisfactory operation of the engineered safeguards systems. 115-kV Temporary line No consequence; 115-kV transmission fault. bus remains energized lines via other line. 115-kV line Breaker fails to The breaker at the breakers trip under line remote terminal of the fault condition. faulted line and the local breaker of the unfaulted line trip to de-energize the 115-kV bus. The local breaker will automatically reclose. If this fails, the 115-kV bus motor-operated NMP Unit 1 UFSAR Section IX IX-6 Rev. 25, October 2017 disconnect switch (8106) will be automatically opened, sectionalizing the bus. The local line breaker (either R40 or R10) will again automatically reclose, energizing a bus section to make offsite power available to one of the redundant engineered safeguards systems at the Operator's discretion. 115-kV line Loss of trip coil, No consequence, since breaker battery power, or each breaker has fully control cable. redundant systems. 115-kV line Low/loss-of-air Low air pressure will breaker pressure. actuate alarm in the control room. Very low pressure will block closing operation. 115-kV line Loss of ac power to Undervoltage relay breaker air compressors. actuates annunciator in control room. Breaker has enough air stored for five closing operations. No air is required for tripping. (Stored Energy Trip) 115-kV bus Bus fault. The bus protective relays will trip both line breakers (R40 or R10). With a live line and dead bus, each breaker will attempt to reclose sequentially. If the first reclosure is successful, the other breaker will close on synchronism check only. If the first reclosure fails, the other breaker will NMP Unit 1 UFSAR Section IX IX-7 Rev. 25, October 2017 attempt to energize the bus by automatic reclosure. If this fails, the 115-kV bus motor-operated disconnect will be opened automatically. Both breakers will automatically reclose a second time. One of the breakers will thereby energize the unfaulted section of the bus, making offsite power available to one of the redundant engineered safeguards systems at the Operator's discretion. 115-kV bus One 115-kV bus Under this condition, section faulted. one group of 4.16-kV power boards, i.e., 11 and 102, or 12 and 103, do not have power available from the 115-kV system. By opening the transformer motor-operated disconnect (8106), this availability can be reestablished via PB 101 through bypass of normal interlocks under procedural control. Reserve Transformer Transformer protective transformer faulted. relays will trip all necessary breakers and open transformer motor-operated disconnect, isolating the transformer from the system. Line breakers will automatically reclose, energizing the 115-kV bus and making offsite power available to one of the redundant engineered NMP Unit 1 UFSAR Section IX IX-8 Rev. 25, October 2017 safeguards systems at the Operator's discretion. Reserve Transformer faulted By removing the faulted transformer and isolated from transformer secondary system. bus links, power from the 115-kV system can be made available to all 4.16-kV power boards, as described under the effect associated with the malfunction "One 115-kV bus section faulted" listed previously. Reserve Low or high The offsite power transformer voltage. system is supervised by load tap the following changer (LTC) protective relays: a. overvoltage b. undervoltage These protective relays are set to transfer PB 102 and 103 to their respective onsite diesel power systems if the voltage of the offsite system deviates beyond set limits required for the satisfactory operation of the engineered safeguards systems. 4.16-kV Loss of voltage. Engineered safeguards PB 101 systems are no longer associated with this power board. Main Generator trip or Normal supply breakers generator transformer fault. will trip and reserve or normal breakers will close by Station fast automatic service transfer, restoring power to PB 11 NMP Unit 1 UFSAR Section IX IX-9 Rev. 25, October 2017 transformer and 12. The 115-kV reserve bus is not affected. 4.16-kV Loss of voltage. Bus undervoltage relay PB 102 and circuit will 103 automatically trip (except for one core spray pump) the normal supply breaker, motor feeder breakers, and will start the diesel generator. The generator breaker will then automatically close, energizing the bus and furnishing power for sequential start of all safeguards systems pumps and 600-V auxiliaries. All 4.16-kV Feeder fault-- Protective relays will power boards three-phase, sense high current and phase-to-phase, and trip feeder breaker. phase-to-neutral. No auto reclosing is provided since a transient fault in metal-clad switchgear is unlikely. All 4.16-kV Feeder fault-- Supply breaker to PB bus power boards breaker fails to will be tripped trip. selectively. No auto reclosing is provided since a transient fault in metal-clad switchgear is very unlikely. Switches are provided to enable the Operator to turn off the automatic reclosers on the 115-kV and 345-kV line breakers for personnel safety, as necessary. 2.0 Station Distribution System NMP Unit 1 UFSAR Section IX IX-10 Rev. 25, October 2017 The basic arrangement of the auxiliary electrical system and the various loads connected to the different power boards are shown on the one-line diagram, Figure IX-1. In general, auxiliaries smaller than 50 hp are fed from 600-V motor control centers (MCC) which in turn are fed from double-ended metal-enclosed unit substations. Auxiliaries ranging from 50 to 300 hp are fed from the unit substations. Loads 300 hp and greater are fed from the 4160-V metal-clad switchgear. The major part of the power required for Station auxiliaries is supplied by the normal auxiliary transformer which is connected to the main generator output leads. This normal auxiliary power transformer, transformer 10, steps down from 24 kV (the nominal generator output voltage) to 4160 V and supplies two separate buses, PB 11 and 12. The Station auxiliary load, other than that fed from PB 101, is divided approximately equally between PB 11 and 12. PB 101 is supplied from the two reserve transformers 101N or 101S which step down from 115 kV to 4160 V. This bus supplies power to one condensate pump, reactor feed booster pump 12, one of five reactor recirculating pumps, the Energy Information Center, the motor-driven fire pump and the sewage plant feeder. Transformer 101N supplies PB 11 when transformer 10 is not available, such as during startup, shutdown and refueling. Transformer 101S supplies PB 12 under similar conditions. Transformers 101S and 101N also supply PB 103 and 102, respectively. Automatic transfer of the auxiliaries from the normal to the reserve source is initiated by low voltage on the auxiliary bus, generator trip, or turbine trip. The transfer is initiated immediately by the protective relays, causing a generator trip or turbine trip, with no intentional delay between the trip of the normal supply breaker and closure of the reserve supply breaker. A sequential-type fast bus transfer is provided on PB 11 and 12 to trip the normal supply breaker and close the reserve supply breaker. In the sequential-type scheme, the reserve supply breaker closes after the normal supply breaker opens (auxiliary "B" contact closes). The power board bus will be disconnected from both sources for approximately 5.2 cycles. NMP Unit 1 UFSAR Section IX IX-11 Rev. 25, October 2017 Low-voltage initiation of transfer is delayed to insure overriding of transient disturbances. In this case, a special relay delays closure of the reserve breaker until the bus voltage has decayed to 20 percent of normal. This is done to preclude reenergizing motors when their residual voltage may be considerably out of phase with the incoming voltage. PB 102 and 103 have as alternate supply sources diesel generators 102 and 103. Automatic transfer of PB 102 and 103 from offsite power to onsite diesel generator power is initiated by low voltage or degraded voltage. PB 102 and 103 feed the post-accident cooling pumps and, through step-down transformers, 600-V PB 16 and 17. PB 16 and 17 feed the low-voltage auxiliaries that are required for Station safety and are vital to safe shutdown under accident conditions. PB 102 feeds the "B" bus section of 600-V PB 16, and PB 103 feeds the "B" bus section of 600-V PB 17. PB 16 and 17 are double-ended, metal-enclosed unit substations, each consisting of two 1,000-kVa step-down transformers and two bus sections, "A" and "B," with incoming and bus tie breakers. The "A" bus section of PB 16 is fed by 4160-V auxiliary feeder 11 from PB 11. The "A" bus section of PB 17 is fed by 4160-V auxiliary feeder 12 from PB 12. The auxiliaries that are required both during normal Station operation and when shut down are duplicated and are connected to the "A" bus sections of PB 16 and 17. If, during shutdown, the 115-kV source is not available to energize PB 11 and 12, the auxiliary feeder secondary breakers will be opened and the bus tie breakers closed with the diesel generators providing the required auxiliary power. The auxiliaries that are required for post-accident systems and for shutdown systems are duplicated and are connected to the "B" bus sections of PB 16 and 17. Power for these auxiliaries is from either the 115-kV system or the diesel generators described later in Section IX-4.1. The 4160-V auxiliary feeders 11 and 12 also feed at the "A" and "C" sections, respectively, of PB 13, 14 and 15. These boards are double-ended, metal-enclosed unit substations, each consisting of two 1,000-kVa step-down transformers and three bus sections, "A", "B" and "C", with incoming and bus tie breakers. NMP Unit 1 UFSAR Section IX IX-12 Rev. 25, October 2017 The Station electrical distribution system is designed with safety and continuity of service as the primary considerations. The Station service transformers, both normal and reserve, are located outdoors with barrier walls between, and are connected to the 4160-V metal-clad switchgear by a nonsegregated phase bus duct. Reserve transformers 101N and 101S are sized, and sufficient facilities are provided, to permit Station startup, shutdown, or operation at reduced load with only one reserve transformer available and the normal transformer OOS. The transformers have automatic load tap changing (LTC) mechanisms on the low-voltage (4160-V) winding. The LTC mechanism is set to automatically maintain voltage on the 4160-V side within a set bandwidth under varying offsite voltage and transformer loading conditions. Manual operation of the LTC mechanism is also controlled from the main control room. Each transformer also has a no-load tap changer on the primary winding. PB 11 and 12 are physically separated for reliability. If either PB 11 or 12 was OOS, the auxiliaries supplied by the other would permit operation at reduced load. PB 101 is located on the floor above PB 11 and 12 for isolation. The low-voltage power boards, 13, 14 and 15, and the MCCs, where feasible, consist of three bus sections with the duplicated auxiliaries on the end sections and the one-of-a-kind auxiliaries on the middle section, which is fed from either end. The 1,000-kVA transformers are all interchangeable, and each is large enough to carry the entire power board load with one transformer disconnected. Either auxiliary feeder 11 or 12 can carry the entire balance of plant low-voltage Station load. The 1,000-kVA transformers themselves are the sealed, dry type and are highly reliable. The substation distribution arrangement is selectively coordinated, using fully-rated breakers with long-time and short-time trip characteristics, to delay the opening of the main circuit breaker until the faulted feeder has had an opportunity to clear. This provides service continuity for all but the faulted circuit. Metal barriers are placed between the bus sections of the metal-enclosed switchgear to confine a bus fault to its own bus section. NMP Unit 1 UFSAR Section IX IX-13 Rev. 25, October 2017 Switchgear other than MCCs use 125-V dc power for control of the electrically-operated circuit breakers. This is so the breakers may be operated if ac power is lost. Dual feeds are provided to the dc control bus on each power board for added reliability, one each from either battery 11, 12 or 14. All equipment that is essential to safety is duplicated, and each duplicated auxiliary system has a cable route for power, control and instrumentation separate and isolated from its counterpart. Many auxiliaries not essential to safety are also duplicated for added reliability and, where possible, have separate cable routes. Included within the scope of the Station auxiliary electrical system are a number of special purpose, limited capacity power sources. These sources are to supply power to the various instrumentation and control, surveillance, computer and alarm systems which are required for both normal and emergency Station conditions. The design basis for each supply, exclusive of obvious voltage, current and quality requirements, as dictated by the nature of the connected load, is mainly concerned with the degree of external electrical system isolation, both long and short term, which it requires because of its associated load. As such, the power sources range from those which are essentially isolated from any external system influence for an extended period of time (hours) to one which is more or less directly coupled to the external system. These supplies, shown on Figure IX-2, include the following. 2.1 Two +/-24-V Dc Systems Each 24-V dc system consists of two 24-V battery chargers (plus and minus) and one 48-V, center-tapped, grounded neutral battery. Each system is intended to supply electrical energy to its assigned part of the source and intermediate range neutron monitoring systems as well as to certain process radiation monitors. When external system power is available, the chargers actually supply the required energy to the connected loads as well as to maintain the battery in a fully-charged condition. Should, for any reason, some or all of the charging capability become unavailable, the battery will inherently and automatically supply the connected load for a minimum of 4 hr. NMP Unit 1 UFSAR Section IX IX-14 Rev. 25, October 2017 2.2 Two 120-V, 60-Hz, Single-Phase, Uninterruptible Power Supply Systems Each reactor protection system (RPS) is supplied power by an uninterruptible power supply (UPS) system. Each UPS system is made up of two UPSs. Each UPS alone is capable of providing power to one RPS power panel. This provides redundant power supplies for each RPS channel. A manual mechanical make-before-break transfer switch selects which one of the two UPSs in each channel will provide power to the associated RPS power panel. In addition, each UPS has an automatic static transfer switch at its output which will transfer to a bypass power source upon sensing either a UPS failure or a downstream high current fault. The bypass power supply is supplied from the 575-V ac input to the UPS, and is conditioned by a step-down isolation transformer with no-load taps. This supply will have a high enough short circuit current capability to permit downstream current faults to be rapidly cleared by protective devices. During normal operation, two UPSs in each channel are energized from the ac supply but only one UPS is connected to the loads by the mechanical transfer switch. The connected UPS is normally powered by 575 V ac from PB 16B or 17B. Should the 575-V ac input be interrupted or the UPS rectifier fail, 120-V ac output power from the UPS inverter will continue to be supplied by the 125-V dc connection to the Station batteries. Should the 125-V dc supply be interrupted or the inverter fail, the static switch internal to the UPS will transfer the 120-V ac loads to the bypass power supply step-down transformer. A sequence of multiple failures and power losses would be required to cause a loss of power to the RPS loads. This is considered to be highly unlikely, since it would require the simultaneous failure of at least two components or power sources. The only single component failures within the UPS system that could interrupt power to the RPS loads is failure of either the static switch or the mechanical transfer switch. Should a sequence of events occur that cause interruption of the 120-V ac UPS output to the RPS loads, plant Operators will be informed by receipt of both a UPS failure alarm annunciator and receipt of those alarms caused by loss of a RPS bus. Operators would then use the mechanical transfer switch to switch the RPS load to the second UPS. The redundant channel UPS is also available when maintenance is required for the in-service UPS. NMP Unit 1 UFSAR Section IX IX-15 Rev. 25, October 2017 This system is extremely reliable because of the many redundant components, multiple power sources, and multiple power paths. Both the ac power source and the dc battery power source can be fed from the Station diesel generators. Should charging to the dc battery bus not be available from the diesel generators (a very unlikely situation), operation of the UPSs will continue on the Station batteries, but will be limited by their capacity. Each UPS system is used to provide power to one of the two RPS buses. Loads on one or both of these buses include power range neutron flux monitoring, RPS instrumentation, unit control, control rod position indication, feedwater instrumentation, some process radiation monitoring, emergency cooling instrumentation, the rod worth minimizer (RWM), Station paging and alarm system, reactor recirculation system motor generator (MG) set controls, etc. 2.3 Two 120-V, 57-60 Hz, One-Phase, Reactor Trip Power Supplies Each reactor trip power supply consists of a two-unit MG set, additional inertia for storing mechanical energy, and necessary control and regulating circuitry. Each of these power supplies provides electrical energy for the loads associated with its respective reactor trip bus. The loads themselves consist mainly of reactor trip solenoid valves and reactor trip contactors, one complete set per reactor trip bus. The stored mechanical energy on each supply prevents spurious trips during a transfer of Station auxiliaries from the unit transformer to the reserve transformer(s), as previously discussed. 2.4 One 120/208-V, 60-Hz, Instrument and Control Transformer The instrument and control transformer is the supply for its associated instrument and control buses (130 and 130A). Normal power supply to the transformer is from PB 13. Loads on bus 130 include miscellaneous chart drives, area radiation monitoring, waste disposal instrumentation, etc. Maintenance supplies from bus 130A are provided for the loads regularly energized from the static UPS and reactor trip power supplies. These alternates are used only under strict administrative control. A switching provision is available for energizing the loads associated with instrumentation and control bus from a 120/208-V supply established from PB 167. 2.5 One 120/240-V, 60-Hz, Single-Phase, Computer Power Supply NMP Unit 1 UFSAR Section IX IX-16 Rev. 25, October 2017 The computer power supply unit is a rotating machinery-type system UPS consisting mainly of a three-unit MG set; an ac (single-phase, 120/240-V) synchronous generator, an ac (three-phase, 550-V) nonexcited synchronous motor, and an appropriate dc machine (125 V). There is also additional shaft-mounted inertia (flywheel), as well as auxiliary regulating and control circuitry, located in two associated freestanding enclosures. During normal operation, each system will operate with its synchronous motor as the driving unit and the alternator supplying its associated connected load. Transfer from normal (ac) to reserve (dc) operation can be initiated by either abnormal voltage or frequency conditions on the external system. The transfer will be automatic and will occur in such a manner that the alternator will smoothly continue to supply the load. Operation from the dc machine and its associated 125-V dc battery will be practically unlimited in time as long as there is dc charging. 3.0 Cables and Cable Trays The basic design criteria for cable runs in trays or conduit include precautionary measures for prevention of cable fires, containment of any fire to a confined area, and protection of cable against fire (other than electrically induced) in hazardous Station areas. The specific design areas are: 3.1 Cable Separation Cables are generally separated functionally, i.e., 4160-V power cable separate from 600-V power cable for major loads, which in turn are separated from low-current 600-V power cables and low-voltage control cables. Cables installed for fire detection and control are separate from all other cables within the plant to prevent the effects of fires, short circuits, etc., from spreading to this system. Vertical fire stops are generally provided where open cable risers occur between floors outside the control room/auxiliary control room and inside the reactor building. Reactor protection and engineered safeguards equipment cables are routed to provide sufficient isolation between similar, functionally-duplicated devices so as to prevent damage in the event of fire or any design basis accident (DBA). NMP Unit 1 UFSAR Section IX IX-17 Rev. 25, October 2017 3.2 Cable Penetrations Cable penetrations associated with primary containment, pressurized boundaries (reactor building and control room), and rated fire barriers have been designed and sealed to prevent a fire from spreading through the penetration. Drywell and suppression pool penetrations are double-sealed, 12-in pipes that are inerted with nitrogen. Pressurized boundary penetrations and rated fire barrier penetrations are sealed using approved sealing details (see Figures VI-2 and VI-3 for description of drywell penetrations). 3.3 Protection in Hazardous Areas Cables contained in cable trays which run through hazardous areas (potential oil fire or explosion) are protected mechanically by the tray itself, by appropriate suppression systems, and/or are coated externally with fire-retardant material for fire resistance. Cables in trays which run past MCCs, power boards, and other equipment that contains material that will support a fire, are generally coated with externally-applied fire-resistant material (Flamemastic), or are protected by design without applying fire-resistant material. Tray shields over the top tray (raised to allow normal ventilation) are provided at selected locations where an opening occurs above the tray to prevent hot weld material or missile damage from above. 3.4 Types of Cables Originally-installed cable construction consists of the following types. 3.4.1 Power Cable 1. 24-kV cable is Kerite insulated with neoprene jackets. 2. 4160-V cable is Kerite insulated with neoprene jackets, or ethylene propylene rubber insulated with Hypalon jackets. 3. Cables for 600-V service and below are cross-linked polyethylene insulated or ethylene propylene rubber NMP Unit 1 UFSAR Section IX IX-18 Rev. 25, October 2017 insulated. Multiple conductor jackets are either polyvinyl chloride (PVC), neoprene or Hypalon. 3.4.2 Control Cable 1. General use, multiconductor control cables are cross-linked polyethylene insulated and 1000-V rated (except supervisory control cable, which is 600-V rated), with Mylar binder tape and either PVC jacket overall or neoprene jacket overall. 2. Special control cables for control rod drive (CRD) instrumentation, area radiation monitors, etc., are multiconductor, PVC insulated, and coordinated voltage rated, with suitable binder tape and PVC jacket overall. 3. Single-conductor control cable inside drywell penetrations (AWG #8, #10, #12 and #16) is cross-linked, polyethylene insulated, type SIS switchboard wire. 3.4.3 Special Cable 1. Coaxial and triaxial cables generally are either solid polyethylene insulated with polyethylene jackets, ethylene propylene rubber insulated with chlorosulfonated polyethylene jackets, or cross-linked polyethylene insulated with cross-linked polyolefin jackets. 2. Shielded cables generally are either polyethylene insulated with PVC jackets, cross-linked polyethylene insulated with chlorosulfonated polyethylene or neoprene jackets, or ethylene propylene rubber insulated with chlorosulfonated polyethylene jackets. Shields are aluminum Mylar tape with an overall covering of flame-retardant tape. 3. Thermocouple extension cable wire generally is PVC insulated and jacketed or cross-linked polyolefin insulated with a Hypalon jacket. Thermocouple extension cables in the drywell are magnesium-oxide insulated with a 304 stainless steel sheath for high temperature. NMP Unit 1 UFSAR Section IX IX-19 Rev. 25, October 2017 4. Substitution for the described special cables may be necessary if the specified cable is not available at the time of replacement or addition. The insulation associated with safety-related cables purchased and installed since the middle of 1974 meets the requirements of IEEE-383 flame test. The insulation associated with nonsafety-related cables purchased and installed since the middle of 1974 also generally meets the requirements of IEEE-383 flame test, except those routed totally in conduit. 3.5 Design and Spacing of Cable Trays 3.5.1 Tray Design Specifications 1. Ladder Tray - Medium-steel cable ladder, 6-, 12- or 24-in wide, typically having 3-in or 4 1/2-in side members of 12-gauge steel and 1-in O.D. rungs spaced on 9-in centers, cold-swaged or welded into side members. System is furnished hot-dip-galvanized after fabrication and used for power cable up to 5 kV and control cable throughout the Station. 2. Solid Tray - Radio-Frequency Communications Tray - Solid 12-gauge steel all around, 24 in wide by 3 in deep, with solid cover and special barriers designed to provide a magnetic path through the cover, dividing the tray into three sections (signal, control, and control power). 3.5.2 Tray Spacing Figures IX-3, IX-4, and IX-5 show the cable routing in the turbine building and typical spacing of all Station cable trays. The vertical spacing shown, e.g., 12 in, is a general spacing only and is not a design criteria. 4.0 Emergency Power 4.1 Diesel Generator System Two sources of electrical power, which are completely separate and are self-contained within the Station and therefore not dependent on any outside source, have been provided by installing two diesel generators. These standby generators each have adequate capacity to start and carry all of the loads required during a maximum emergency power requirement period. NMP Unit 1 UFSAR Section IX IX-20 Rev. 25, October 2017 Each is designed to be capable of starting and picking up load in 10 sec or less, when lube oil and jacket water temperatures are at or above 85°F and ambient temperature at least 50°F. These diesel generators will start automatically on loss of voltage at the bus to which they are connected. A manual start control with dead line pickup, manual synchronizing or automatic synchronizing, is also provided. Power to start each unit using air motors for cranking is provided by an air system consisting of five air tanks and two compressors. Each air system stores sufficient air for five starts. There are no ties between air systems. Should a unit fail to start on signal, it is automatically programmed for a second attempt. Dc control power for starting is available from two completely separate Station batteries. One battery normally supplies each diesel generator control circuit with provision to obtain control power from the other battery by manual transfer switching if necessary. The fuel supply for these diesel generators is contained in two 12,000-gal storage tanks, which are cross-connected (outside the diesel generator building) so that either unit can take oil from either tank, but each tank is normally assigned to one unit and is valved off from the other. When connected in parallel, these tanks have the capacity to permit one diesel unit to operate for 4 days at full load. By utilizing normal commercial deliveries of diesel fuel, either or both units can be operated continuously. The diesel generator fuel oil system is shown on Figure IX-8. The quality of the fuel oil is monitored by the periodic sampling of the emergency diesel generator storage tank, through the tank fill port, in accordance with approved industry standards. These samples are analyzed for parameters indicative of fuel oil degradation and contamination impurities. The tanks are double walled to permit leak testing in accordance with Department of Environmental Conservation (DEC) and Environmental Protection Agency (EPA) regulations. In addition, a concrete-paved spill pad around the fill points and piping to a spill collection tank were added to control delivery spills. The nameplate rating of each diesel generator is 3125 kVA at a 0.8 p.f. (power factor). Based on a calculated p.f. of 0.86, each diesel generator is rated for continuous operation of 2586 kW (3125 kVA). This rating is limited by the engine output. The synchronous generator is rated at 3125 kVA continuous and 3440 kVA for overload ratings. The diesel generator is also NMP Unit 1 UFSAR Section IX IX-21 Rev. 25, October 2017 rated for a 10-percent overload of 2845 kW allowable for 2 hr in a 24-hr period (limited by engine output). In addition to the above, each diesel generator has a 2000 hr/yr rating of 2838 kW and a 7-day/yr emergency rating of 2945 kW (again limited by engine output) at the expense of increased maintenance. The important factor is not maintenance, but the availability and ability of these units to carry an emergency load. Therefore, it can be safely assumed that for the use intended, each unit may be loaded to 2945 kW. Maximum anticipated continuous load, as determined by analysis, is maintained within the 2000 hr/yr rating. Figure IX-1 shows the electrical one-line diagram of the two diesel generator systems. To protect the generators from damage which could occur if they were connected to the system while out of phase, the circuit breakers in the feeders from PB 101 are electrically interlocked so that they cannot be closed if the diesel generator breaker to the same power board is already closed. However, to allow for testing the unit fully loaded, the generator breaker can be closed on a live bus if the unit is either manually or automatically synchronized. Each generator is protected from faults by the following relays: 1. Generator induction time overcurrent relay. 2. Neutral induction time overcurrent relay. 3. Overvoltage relay. 4. Generator differential relay. 5. Field ground relay. 6. Field overcurrent relay. 7. Loss of excitation relay. 8. Instantaneous directional ground relay. Each diesel engine is protected by the following devices: 1. Engine temperature switch. 2. Bearing oil pressure switch. NMP Unit 1 UFSAR Section IX IX-22 Rev. 25, October 2017 3. Overspeed switch. 4. Low water pressure. 5. Pressurized crankcase low pressure. 6. Dc motor-driven prelube oil pumps. To facilitate start of the diesel generators, all 4160-V motor breakers and all nonessential 600-V loads are tripped automatically upon loss of bus voltage except for one core spray pump. The essential 600-V loads are picked up when the generator breaker closes and the necessary 4160-V motors are started automatically, in sequence. Sequence starting of these large motors is necessary to prevent momentary overloading of the unit and excessive voltage dip, which would result from the large starting kVA requirement if all of these motors started simultaneously. The loads supplied by the diesel generators are grouped into two main categories: 1. Loads required following a loss-of-coolant incident. 2. Loads required for orderly shutdown. Item 1 above can be further subdivided into safety system loads which are started automatically, and other less critical loads which are started manually. The maximum emergency power requirement would occur for a loss-of-coolant incident. Figures IX-6a and IX-6b show the sequence of loads on the diesel generators for a loss-of-coolant accident (LOCA). Figure IX-7 shows the sequence of loads on the diesel generators for an orderly shutdown. Each diesel generator is also available on a manual basis for other loads in addition to the required automatic loads. The connection to such other loads is made with regard to overload restrictions under administrative procedural control. The primary basis for selecting a standby generator is that it must be truly independent. Actually, normal sources of power are very reliable and the probability of truly random coincident NMP Unit 1 UFSAR Section IX IX-23 Rev. 25, October 2017 failures of all sources of power into the Station is very low. If the failures were truly random, standby power would probably not be necessary. However, all the external sources of power entering the Station are carried on overhead lines with a certain vulnerability to storms of wind, lightning and ice. The standby diesel generators are provided to guard against the contingency of the concurrent forced outage of all external emergency sources of power, and it is imperative that these generators not be influenced by the same environments that affect the normal sources of power. For these reasons, the diesel generator units are not connected, aside from fuel, with each other or with any outside power sources. The diesel generators and PB 102 and 103 are located at grade elevation in rooms connected to the turbine building. Each power board and each diesel generator are located in a separate room so that an incident at one unit should not involve the other. 4.2 Station Batteries Two separate and independent Station batteries are provided. Each is a nominal 125-V, 2320-amp/hr battery of the lead-calcium type. Conventionally, the Station battery is used as a power source for breaker operation and instrumentation and control because of its proven reliability and availability during a momentary or prolonged loss of ac power. Where loss of ac power, even for a short time, could result in serious equipment damage and excessive maintenance, emergency systems are often powered by the Station battery. For this generating Station, the heavy-duty, short-duration loads have determined the required capacity of the storage batteries. Having been established, the requirement is met by providing two independent 125-V, 2320-amp/hr batteries, as shown on Figure IX-2. This philosophy of a dual dc system is carried throughout the Station. Each battery feeds into its own battery board. There are no cross-connections, thus eliminating any possibility of inadvertently interconnecting the two dc systems. Each battery is kept charged by two static chargers that maintain the battery floating fully charged. Although the battery chargers can be fed from the normal Station auxiliary system, their usual source of power is from the Station reserve supply. This eliminates the transfers and momentary outages NMP Unit 1 UFSAR Section IX IX-24 Rev. 25, October 2017 which occur every time the main unit is taken out of service. In an emergency, the chargers will be supplied from diesel generators. Each charger is sized to supply the dc power required by steady-state loads (steady-state load equals continuous plus noncontinuous loads, but not transient loads), and also to restore the battery from design minimum to the fully-charged state within a limited time span (8-12 hr), regardless of the status of the Station. Additionally, the chargers provide adequate voltage to float and equalize charge the batteries. Mentioned elsewhere in this section are two static UPS systems for channels 11 and 12 that provide power for the dual-logic protective system, instrumentation systems and the communication and alarm systems. Under the emergency conditions described above, each functions as an inverter fed from the battery to provide continuous, uninterrupted ac power to specific loads as required. A computer is included with the auxiliary equipment for this generating Station. Continuous power to the computer is desirable since the computer has both sequence of events and postmortem priority interrupt programs. Continuous power supply is assured by using a three-unit MG set 167, which functions as an inverter fed from the battery to provide continuous uninterrupted ac power to the computer. This unit is also available as a spare battery charger if required. Transfer of the dc unit of this MG set to either battery is performed manually through key-interlocked switches at the battery boards and is performed function only. Change of mode of operation from inverter to battery charger is performed manually at the MG set control cubicle. Dual feeds supply the power board control buses, one each from either battery 11, 12 or 14. All high-voltage breakers (345 kV and 115 kV) have dual-trip circuits supplied, one each from either battery 11, 12 or 14. In the event of an accident which requires Station evacuation accompanied by a complete absence of both offsite and onsite ac power, the emergency dc lighting system will provide ample time for evacuation. In view of the redundancy of the emergency power sources, which consist of one 115-kV line from a hydroelectric generating NMP Unit 1 UFSAR Section IX IX-25 Rev. 25, October 2017 station, a separate 115-kV line from a nearby fossil-fueled generating station, and two independent onsite diesel generators, it is extremely unlikely that these Station batteries will ever be called upon for a major power output for longer than a very few minutes during the anticipated life of the Station. Because of the redundancy of emergency power, it is unlikely that a condition will exist where neither diesel generator will start. With one diesel generator, both batteries can be kept under charge. The magnitude and duty cycle of major battery loads following a loss of offsite power (LOOP) were evaluated for two separate cases. Case "a" involves LOOP with maximum Technical Specification leakage, and Case "b" involves LOOP concurrent with a LOCA. Table IX-1 shows the magnitude and duty cycle of major battery loads for the Case "b" event, which is the more bounding of the two events with respect to battery loading. With no ac power, the batteries will meet all reactor safety load requirements in the early hours and provide lighting in the control room area. Each battery is located in a separate ventilated battery room of generous proportions to insure ample room for testing and maintenance. Each battery board is also in a separate room, accessible from either inside or outside the Station. Each dc system is operated independently and ungrounded, with ground detection devices to indicate the first grounds. Thus, with reasonable care, multiple grounds--the only probable mode of failure--are extremely unlikely. Battery cells are mounted on two two-step racks with earthquake bracing. Individual electric heaters, which are thermostatically controlled, maintain the battery rooms at an optimum temperature. 4.3 Q-Related Battery System A Q-related battery system, comprised of one 125-V dc 1552 amp-hr Q-related battery (battery #14) connected to a Q-related battery board (#14), provides power to several high-current loads. This battery is connected to the battery board by a breaker. A nonsafety-related 125-V dc 500 amp static battery charger (#14), supplied by PB 15, is connected to the battery board to provide both charging power and to supply the constant dc load. The major loads on the Q-related battery are the turbine generator dc emergency bearing oil pump (40 hp) and the emergency hydrogen seal oil pump (7-1/2 hp). Neither is NMP Unit 1 UFSAR Section IX IX-26 Rev. 25, October 2017 required when ac power is available. If all ac power fails and does not become available in the interim, both pumps will start automatically soon after the turbine generator trips. The bearing oil pump will run for the one-half hour it will take for the unit to come to rest. The seal oil pump may be run for 8 hr if desired. Neither pump will be required again until conditions are restored to normal. A single seismically-designed masonry enclosure with a steel-decked concrete roof encloses the battery for personnel safety and fire protection. The room is a 1 1/2 hr fire-rated enclosure. The room is ventilated to remove explosive gases (hydrogen). The air turnover rate is sufficient to insure that an explosive mixture will not exist at the maximum hydrogen generation rate. Operators will be alerted to a loss of the exhaust system forced-air fan by an existing fan failure alarm in the main control room. This will provide warning of a potential explosive gas buildup. The room temperature is maintained greater than 65°F by the surrounding heat capacity of the turbine building during normal plant operations. Fire-rated dampers are installed in the ventilation system to isolate the battery room in case of fire. The battery room is located on el 277' between columns G10, G12, F12 and F10. The static battery charger and the battery board are located along the east battery room wall. 5.0 Tests and Inspections 5.1 Diesel Generator Since these are standby units, readiness is of prime importance. Readiness can best be demonstrated by periodic testing. The testing confirms the generator's ability to start the connected load, and is of such duration as to bring everything to equilibrium conditions, assuring that cooling and lubrication are adequate for extended periods of operation. The ability of the standby generator to start rapidly upon demand is consistent with the concept of maintaining continuity of core cooling under accident or emergency conditions. Further, it is considered that a continuous source of auxiliary power is assured for the Station by the facilities described above. The following tests will be included in appropriate Technical Specifications. NMP Unit 1 UFSAR Section IX IX-27 Rev. 25, October 2017 1. At monthly intervals during normal operation, the diesel generators are started, paralleled with the system, and operated for a minimum of 1 hr at rated load to assure that all systems are functioning properly. The monthly testing is based on the manufacturer's recommendation for these units in this type of service. 2. During each refueling outage, the diesel generators are tested for automatic start and automatic pickup of all loads which would be required if a loss-of-coolant condition existed. 3. In addition, visual checks and inspections at specified periods, as recommended by the manufacturer, are made. These include inspection for leaks, checking fluid levels, lubrication, replacement of air and oil filters, etc. 4. The diesel cooling raw water pumps are tested quarterly in accordance with the Unit 1 In-Service Testing (IST) Program. 5.2 Station Batteries Monthly inspections are made to determine the specific gravity of each cell, as well as the battery voltage. Inspections on a weekly basis are also made to determine the individual cell voltage and the specific gravity of the pilot cells of each battery. 5.3 Q-Related Battery Periodic inspections are made to determine the specific gravity of each cell, as well as the battery voltage. Inspections are also made to determine the individual cell voltage and the specific gravity of the pilot cell of the battery. 6.0 Conformance With 10CFR50.63 - Station Blackout Rule Station blackout (SBO) is defined in 10CFR50.2. The SBO rule requires that each light-water-cooled nuclear power plant be able to withstand and recover from a SBO of a specified duration. Regulatory Guide (RG) 1.155 described a method acceptable to the Nuclear Regulatory Commission (NRC) staff for meeting the requirements of 10CFR50.63. NMP Unit 1 UFSAR Section IX IX-28 Rev. 25, October 2017 NUMARC 87-00, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," also provides guidance that is in large part identical to the RG 1.155 guidance and is acceptable to the NRC staff for meeting these requirements. Table 1 of RG 1.155 provides a cross-reference between the regulatory guide and NUMARC 87-00, and notes where the regulatory guide takes precedence. Unit 1 has been evaluated against the requirements of the SBO rule using guidance from NUMARC 87-00, and the supplemental guidance provided by NUMARC 87-00, Supplemental Questions/Answers, December 27, 1989, and NUMARC 87-00, Major Assumptions, December 27, 1989, except where RG 1.155 takes precedence. The results of this evaluation were submitted to the NRC in References 1, 2, 3, 5, 6 and 8 and are summarized below. NRC evaluations and acceptance of the Unit 1 response to the SBO rule were documented in References 4, 7 and 9. 6.1 Station Blackout Duration A SBO duration of 4 hr was determined based on the following plant factors: 1. Ac Power Design Characteristic Group is "P2" based on: a. Expected frequency of grid-related LOOP events does not exceed once per 20 yr. b. Estimated frequency of LOOP due to extremely severe weather places the plant in extremely severe weather group 1. c. Estimated frequency of LOOP due to severe weather places the plant in severe weather group 3. d. The offsite power site is in the I3 group. e. Plant-specific prehurricane shutdown requirements and procedures are not required for Unit 1, nor are such procedures credited in the determination of the Ac Power Design Characteristic Group. 2. The Emergency Ac Power Configuration Group is "C" based on: NMP Unit 1 UFSAR Section IX IX-29 Rev. 25, October 2017 a. There are two emergency ac power supplies not credited as alternate ac power sources. b. One emergency ac power supply is necessary to operate safe shutdown equipment following a LOOP. 3. The target emergency diesel generator reliability is 0.975. A target emergency diesel generator reliability of 0.975 was selected based on having a nuclear unit average emergency diesel generator reliability for the last 20 demands greater than 0.90. An analysis showing the emergency diesel generator reliability statistics for the last 20, 50, and 100 demands which supports this target reliability has also been performed. 6.2 Station Blackout Coping Capability The characteristics of the following plant systems were reviewed to assure that the systems have the availability, adequacy and capability to achieve and maintain a safe plant shutdown and to recover from a SBO for the 4-hr coping duration. Condensate Inventory for Decay Heat Removal It has been determined that 58,700 gal of water are required for decay heat removal and cooldown for 4 hr. The minimum permissible emergency condenser (EC) gravity feed, EC makeup tank and EC levels, per Technical Specifications, provide 114,720 gal of water, which is adequate to provide for decay heat removal for at least 4 hr even if both EC level control valves fail open on loss of air at the start of the SBO, and no other Operator actions are taken. Therefore, no plant modifications or Operator actions are required to ensure adequate condensate capacity exists for decay heat removal during a 4-hr SBO. The design basis SBO calculations note that if Operators secure one EC from service to control vessel cooldown rate, then manual actions are required within 30 min to conserve condensate inventory and maximize coping duration. The calculations show that the actions to isolate the idle EC from the makeup tank, open the crosstie, and take manual control within 30 min, is required to minimize any overflow from the ECs into the waste building. These actions within 30 min will ensure a coping period of 4.3 hr. NMP Unit 1 UFSAR Section IX IX-30 Rev. 25, October 2017 The SBO analysis includes a case that demonstrates the cooldown is below the design analysis 300° per hr emergency cooldown rate and, as such, Operator action to secure one EC is not required. Therefore, securing one EC is not a requirement of the SBO analysis. The decision/option to secure one EC based on cooldown rate is at the discretion of the Operator in accordance with EOPs. Station Battery Capacity Battery capacity calculations performed pursuant to NUMARC 87-00, Section 7.2.2, and IEEE-485-1978, verified that the Station batteries have sufficient capacity to meet SBO loads for 4 hr. Operator action is required to shed nonessential loads from Class 1E batteries to cope with a SBO duration of 4 hr. The shedding of the nonessential loads from Class 1E batteries is identified in plant procedures. Compressed Air Air-operated valves (AOVs) relied upon to cope with a SBO for 4 hr can either be operated manually or have sufficient backup sources independent of the preferred and blacked out Unit's Class 1E power supply. Valves requiring manual operation or that need backup sources for operation are identified in plant procedures. Effects of Loss of Ventilation The key areas in which the loss of ventilation cooling causes a concern for equipment operability were identified based on the equipment used to respond to the SBO event. Heatup calculations were performed for the: 1. EC condensate return isolation valve room (el 281') 2. EC steam supply isolation valve room (el 298') 3. Reactor building, el 318' 4. Reactor building, el 340' 5. Primary containment 6. Control room NMP Unit 1 UFSAR Section IX IX-31 Rev. 25, October 2017 The control room at Unit 1 does not exceed 120°F during a SBO and, therefore, is not a dominant area of concern (DAC). Reasonable assurance of the operability of SBO response equipment in the dominant areas of concern has been assessed using Appendix F to NUMARC 87-00 and the Topical Report. No hardware modifications are required to provide reasonable assurance for equipment operability. Procedures direct the Operators to open the control room and auxiliary control room instrument cabinet doors which will increase the cooling of the control room equipment by natural convection. Containment Isolation The plant list of containment isolation valves has been reviewed to verify that valves which must be capable of being closed or that must be operated (cycled) under SBO conditions can be positioned (with indication) independent of the preferred and blacked-out Class 1E power supplies. Plant procedures identify valves which must be operated to isolate containment during a SBO. Reactor Coolant Inventory An analysis of reactor coolant system (RCS) inventory was performed assuming a leak rate of 18 gpm per recirculation pump (5 pumps) and the maximum allowable (25 gpm) Technical Specification leak rate. The results indicate that reactor water level would reach top of active fuel (TAF) in approximately 1.8 hr. With a constant leak rate of 115 gpm, plant procedures direct the Operator to actuate the automatic depressurization system (ADS) at or before the time the water level reaches the minimum steam cooling RPV water level (MSCRWL). After the vessel is depressurized, plant procedures direct the Operator to initiate reactor vessel makeup using the diesel-driven fire pump. 6.3 Procedures and Training Plant procedures, SBO response guidelines, ac power restoration procedures, and SW procedures have been reviewed, and changes necessary to meet NUMARC 87-00, Section 4, guidelines have been implemented to ensure an appropriate response to a SBO event. NMP Unit 1 UFSAR Section IX IX-32 Rev. 25, October 2017 Personnel training to ensure an effective response to a SBO event has been incorporated into the training program. 6.4 Quality Assurance Based on a review of the equipment relied upon to carry out the SBO response, all nonsafety-related components have been upgraded to a "Q" classification and are covered under the Quality Related Program for Nine Mile Point Nuclear Station Operations, which is consistent with the guidance of RG 1.155, Appendix A. The remaining SBO equipment is safety related and is covered by existing quality assurance requirements in the Quality Assurance Topical Report (QATR). 6.5 Emergency Diesel Generator Reliability Program An Emergency Diesel Generator Reliability Program has been developed for Unit 1 which conforms to the guidance of RG 1.155, Position C.1.2. The program includes a 0.975 emergency diesel generator target reliability based on emergency diesel generator reliability data for the last 20, 50 and 100 demands. 6.6 References 1. Letter NMP1L 0384 from C. D. Terry (NMPC) to NRC, dated April 13, 1989. 2. Letter NMP1L 0489 from C. D. Terry (NMPC) to NRC, dated April 3, 1990. 3. Letter NMP1L 0491 from C. D. Terry (NMPC) to NRC, dated April 16, 1990. 4. Letter from D. S. Brinkman (NRC) to L. Burkhardt, III (NMPC), Request for Additional Information, Nine Mile Point Unit 1 (TAC No. 68570), dated November 20, 1990. 5. Letter NMP1L 0558 from C. D. Terry (NMPC) to NRC, dated January 7, 1991. 6. Letter NMP1L 0564 from C. D. Terry (NMPC) to NRC, dated January 24, 1991. 7. Letter from D. S. Brinkman (NRC) to B. R. Sylvia (NMPC), Station Blackout Rule Safety Evaluation - Nine Mile Point NMP Unit 1 UFSAR Section IX IX-33 Rev. 25, October 2017 Nuclear Station Unit No. 1 (TAC No. 68570), dated July 1, 1991. 8. Letter NMP1L 0599 from C. D. Terry (NMPC) to NRC, dated August 1, 1991. 9. Letter from D. S. Brinkman (NRC) to B. R. Sylvia (NMPC), Station Blackout Rule Supplemental Safety Evaluation - Nine Mile Point Nuclear Station Unit No. 1 (TAC No. 68570), dated November 6, 1991. NMP Unit 1 UFSAR Section IX IX-34 Rev. 25, October 2017 TABLE IX-1 MAGNITUDE AND DUTY CYCLE OF MAJOR STATION BATTERY LOADS Loads (amps) Case "b" 0-1 Minute (amps) 1-2 Minutes (amps) Battery 11 Battery Board 11 Instrument and Control Power UPS System 162 Other Continuous Loads Diesel Generator 102 Start and Field Flashing, Run 211 103 60 211 82 34 Breaker Trips Two - 345 kV Two - 115 kV Twelve - 4160-V PB 11 Three - 4160-V PB 101 One - 4160-V R1012 Six - 600-V PB 16 One - 500-V dc Generator Field Bkr 21.0 14.9 72.0 18.0 6.0 12.0 6.0 Breaker Closures One - 4160-V PB 102 (Diesel Generator) Six - 4160-V PB 102 (ECCS Equip.) One - 600-V PB 16 (CRD, SBC) 14.0 14.0 44.0 44.0 Battery 12 Battery Board 12 Instrument and Control Power UPS System 172 MG Set 167 Other Continuous Loads Diesel Generator 103 Start and Field Flashing, Run 211 143 87 60 211 143 69 34 NMP Unit 1 UFSAR Section IX IX-35 Rev. 25, October 2017 TABLE IX-1 (Cont'd.) Loads (amps) Case "b" 0-1 Minute (amps) 1-2 Minutes (amps) Breaker Trips Eleven - 4160-V PB 12 One - 4160-V R1014 One - 4160-V R1013 One - 600-V PB 14 Five - 600-V PB 17 66.0 6.0 6.0 2.0 10.0 Breaker Closures One - 4160-V PB 103 (Diesel Generator) Six - 4160-V PB 103 (ECCS Equip.) Two - 600-V PB 17 (CRD, SBC) 14.0 14.0 44.0 44.0 Motor-Operated Valves Reactor Cleanup Supply IV #12 (33-04) 106 NOTES: 1. Case "b" assumes loss of 115-kV offsite power combined with LOCA and unit trip. Case "a" assumes loss of 115-kV offsite power combined with Technical Specification leakage, but without a unit trip. The Case "b" event bounds the Case "a" event with respect to battery loading. Therefore, the Case "a" event is not included in the Table. 2. Continuous and noncontinuous loads are supplied from the battery until the static battery charger is transferred to its ac power source. NMP Unit 1 UFSAR Section IX IX-36 Rev. 25, October 2017 TABLE IX-1 (Cont'd.) 3. ECCS breaker closures are staggered utilizing time delays. There are no closure overlaps; therefore, the breaker close current for one breaker is seen throughout the first minute. For detailed load and duty cycle information, see the battery sizing calculations.

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N t TRAYS BELOW EL.277' i: .--------1 t j TURBINE REACTOR I I EXTENSION BUILDING . I ! BUILDING WALL l SLEEVES I: -r---_-_-_--_____ 4_-_-_-_-1-I ----------------------------L---------. I I I *-3-24* TRAYS <CONTROL> *---3-24' TRAYS CPOWER & CONTROL> J...a ..JL I I I I t BUILDING t 14', ,411 r: n I :---------------------------------------1 I L ______________ _J A = LADDER TRAY 120v o-o 125v d-c B = b-ftDDER TRAY 31p.se0v C = LADDER TRAY 120v ari-o 125v d-c D = LADDER TRAY l-12v d-c <28v o-c FIGURE IX-4 UFSAR Rev. 14 (June 1996)

• 24' TRAYS <CONTROL> TRAYS BELOW EL.300' N t REACTOR BUILDING WALL
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• 321KW CRO PUMP 11 -243Kll MANUAL START, QR 112 NOTE: ACTUAL KW VALUES s;. THOSE SHOWN. REFER TO THE O!ESEL GENERATOR LOADING CALCULATION FOR THE ACTUAL DIESEL GENERATOR LOAD VALUES. THIS SHEET DERIVES FROM CRP* Hl'HJFS-lill7. _______ A_ .fil§19!_ ________________________________________ _ e 0 2 4 6 8 18 IZ 11 16 18 2B 2Z 24 26 28 J8 32 J< 36 38 12 1 14 46 18 5:1 52 51 56 58 lll 11 l2 13 15 Z 3 1 13 2s I __ HOURS-----j TIME FIGURE IX-6a UFSAR Rev1s1on 24 October 2015 3 I 0 <! 0 ...J 3'.-,... -3211 Jiii -... .... .... -2509 *** .,. ... ... 2eee *-... ,,. *-15'8 ---u* 1ee0 --,. .. 501 ---** DIESEL GENERATOR #103 LOADING FOLLOWING LOSS-OF-COOLANT ACCIDENT m" Pifi\ 1Z2 121 Si). 122 RH> 112 PIH' 12 i-Y.!9,.RESIART OF m"l'OISllllPUHPl2 MANUAL iRIP OF: CCJllE SPRAY TOPP!NG PUMP
• 321KW CRD PUMP 12
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• 112KN -& !iffuism -HKW lj6'1P ..JI.cw"" P 121 OA 122 NOTE: ACTUAL KW VALUES ARE :iO: THOSE SHOWN. REFER TO THE DIESEL GENERATOR LOADING CALCULATION FOR THE ACTUAL DIESEL GENERATOR LOAD VALUES. THIS SHEET DERIVES FROM CRP* l-0q-uFS-017. _______ . ..ll!§&.l§L !§ZIL&...!t !.._Cefil.UCl§§*_ll§!!' _________________________________________ _ e 4 111214118 3&4144446481 I 2 1 13 I 14---------------SECOllDS MIMUTES HOURS----J TIME FIGURE IX-6b UFSAR Rev1s1on 23 October 2013 18 z 5! !!! > II! I .. i§ ; ffi

( I I fSAR CRO 0271.4 DIESEL GENERATOR LOADING FOR ORDERLY SHUTDOWN 1';1.110 R£5Tl<Rl OF CONTROLROOtRi'IEi't.H'll 50 PGWE.R l!Ol'RO 16\B IWJlOR£ll.lŽT0A(f0WffiOF, B1<TTmVC:11Af'IGEF;S-lShVB MA"UAL.':lTARTOf'EMERGE'.'"1:*$ERV!CE lll<<TER !H50t1P lt.ISTRIJMENT"'TICN l'ltl! CO"TRQ FOO/ER I.I'S 162 COMPUTER MG Se:T UP STMT OF, REACTOR COOLJl-lC Pl.!'1F 11 OR ----------------------------------------------------------------------------FIGURE IX-7 UFSAR Flev. 14 (June 1996) . I I I DIESEL GENERATOR FUEL OIL SYSTEM ALT. FILL LEAK DETECTOR n HH/LL Ej---------------------------------0 i A--AUTO SAAT/STOP---1 *-v, : i i i : : : : : I DIESEL GENERATOR 11102 FUEL OILHOAY TANK 41'.ll2l GAL. LE HH/LL 0--------i 0-------------------------------Lj !---AUTO SART . 1V, ' ' ' l ! l ' ' ' ' ' ' ' ' ' : ! ! i ' H/L

• HH/LL DIESEL GENERATOR "103 FUEL Oll..21BJAY TANK 400 GAL. ALT. FILL VENT ALARM --C}v------------EJ--: 0 DETECTOR ' LA --jf-LA l -------LE FIGURE IX-8 UFSAR REVISION 17 OCTOBER 2001 U.S. NUCLEAR REGULATORY COMMISSION DOCKET 50-220 LICENSE DPR-63 NINE MILE POINT NUCLEAR STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT (UPDATED) OCTOBER 2017 REVISION 25 NMP Unit 1 UFSAR LIST OF EFFECTIVE FIGURES SECTION X Figure Revision Number Number Section X EF X-1 Rev. 25, October 2017 X-1 22 X-2 22 X-3 21 X-4 17 X-5 22 X-6 25 X-7 17 X-8 24 X-9 22 X-10 14 X-11 14 NMP Unit 1 UFSAR Section X REACTOR AUXILIARY AND EMERGENCY SYSTEMS TABLE OF CONTENTS Section Title Section X X-i Rev. 25, October 2017 A. REACTOR SHUTDOWN COOLING SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections B. REACTOR CLEANUP SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections C. CONTROL ROD DRIVE HYDRAULIC SYSTEM 1.0 Design Bases 2.0 System Design 2.1 Pumps 2.2 Filters 2.3 First Pressure Stage 2.4 Second Pressure Stage 2.5 Third Pressure Stage 2.6 Exhaust Header 2.7 Accumulator 2.8 Scram Pilot Valves 2.9 Scram Valves 2.10 Scram Dump Volume 2.11 Control Rod Drive Cooling System 2.12 Directional Control and Speed Control Valves 2.13 Rod Insertion and Withdrawal 2.14 Scram Actuation 3.0 System Evaluation 3.1 Normal Withdrawal Speed 3.2 Accidental Multiple Operation 3.3 Scram Reliability 3.4 Operational Reliability 3.5 Alternate Rod Injection 4.0 Reactor Vessel Level Instrumentation Reference Leg Backfill 5.0 Tests and Inspections NMP Unit 1 UFSAR Section X REACTOR AUXILIARY AND EMERGENCY SYSTEMS Section Title Section X X-ii Rev. 25, October 2017 D. REACTOR BUILDING CLOSED LOOP COOLING WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections E. TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections F. SERVICE WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections G. MAKEUP WATER SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections H. SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections I. BREATHING, INSTRUMENT AND SERVICE AIR SYSTEM 1.0 Design Bases 2.0 System Design 3.0 Design Evaluation 4.0 Tests and Inspections NMP Unit 1 UFSAR Section X REACTOR AUXILIARY AND EMERGENCY SYSTEMS Section Title Section X X-iii Rev. 25, October 2017 J. FUEL AND REACTOR COMPONENTS HANDLING SYSTEM 1.0 Design Bases 2.0 System Design 2.1 Description of Facility 2.1.1 Cask Drop Protection System 2.2 Operation of the Facility 2.3 Control of Heavy Loads Program 2.3.1 Introduction/Licensing Background 2.3.2 Safety Basis 2.3.3 Scope of Heavy Load Handling Systems 2.3.4 Control of Heavy Loads Program 2.3.4.1 NMPNS Commitments in Response to NUREG-0612, Phase I Elements 2.3.4.2 Reactor Pressure Vessel Head and Spent Fuel Cask Lifts 2.3.5 Safety Evaluation 3.0 Design Evaluation 4.0 Tests and Inspections K. FIRE PROTECTION PROGRAM 1.0 Design Bases Summary 1.1 Defense-in-Depth 1.2 NFPA 805 Performance Criteria 1.3 Codes of Record 2.0 System Description 2.1 Required Systems 2.2 Structures 3.0 Safety Evaluation 4.0 Fire Protection Program Documentation, Configuration Control and Quality Assurance L. REMOTE SHUTDOWN SYSTEM 1.0 Design Bases 2.0 System Design 3.0 System Evaluation 4.0 Tests and Inspections M. HYDROGEN WATER CHEMISTRY AND NOBLE METAL CHEMICAL ADDITION (NOBLECHEM)

NMP Unit 1 UFSAR Section X REACTOR AUXILIARY AND EMERGENCY SYSTEMS Section Title Section X X-iv Rev. 25, October 2017 SYSTEMS 1.0 Design Basis 1.1 Noble Metal Chemical Addition System 1.2 Hydrogen Water Chemistry System 2.0 System Design 2.1 Noble Metal Chemical Addition 2.2 Hydrogen Water Chemistry System 2.2.1 HWC Feedwater Hydrogen Injection 2.2.2 HWC Offgas Oxygen Injection 2.2.3 HWC Offgas Sample 3.0 System Evaluation 4.0 Tests and Inspections N. REFERENCES NMP Unit 1 UFSAR Section X LIST OF FIGURES Figure Number Title Section X X-v Rev. 25, October 2017 X-1 REACTOR SHUTDOWN COOLING SYSTEM X-2 REACTOR CLEANUP SYSTEM X-3 CONTROL ROD DRIVE HYDRAULIC SYSTEM X-4 REACTOR BUILDING CLOSED LOOP COOLING SYSTEM X-5 TURBINE BUILDING CLOSED LOOP COOLING SYSTEM X-6 SERVICE WATER SYSTEM X-7 FIGURE DELETED X-8 SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM X-9 BREATHING, INSTRUMENT, AND SERVICE AIR X-10 REACTOR REFUELING SYSTEM PICTORIAL X-11 CASK DROP PROTECTION SYSTEM NMP Unit 1 UFSAR Section X X-1 Rev. 25, October 2017 SECTION X REACTOR AUXILIARY AND EMERGENCY SYSTEMS A. REACTOR SHUTDOWN COOLING SYSTEM 1.0 Design Bases This system, shown on Figure X-1, is designed to cool reactor water below temperatures and pressures at which the main condenser may be used as a heat sink following reactor shutdown. Once the reactor water has been cooled to about 350°F by the main condenser, the shutdown cooling system is used to cool the reactor water down to 125°F and maintain it at this temperature by removing fission product decay heat absorbed by the reactor water. 2.0 System Design At or below a reactor pressure of 120 psig and a temperature of 350°F, the shutdown cooling system may be manually actuated from the main control room. The reactor water enters this system from the suction side of one of the reactor recirculation pumps, flows through the partial-capacity shutdown cooling system loops, then discharges into the discharge side of another reactor recirculation loop pump. Heat removal requirements are variable depending on the operating condition prior to shutdown. Usually two heat exchangers and two pumps are sufficient to provide a desirable cooldown rate (100°F/hr maximum allowable). If necessary, all three loops of the shutdown cooling system can be used. The heat exchangers are cooled by demineralized water from the reactor building closed loop cooling water (RBCLCW) system, described in Section X-D. Each of the three heat exchangers is designed for a heat load of 12,500,000 Btu/hr. The equipment in this system is designed and constructed in accordance with ASME Pressure Vessel Code 1965, Section III, Class C, for the parts in contact with reactor water. Piping, valves other than those for isolation, and fittings are built to the requirements of ASA B-31.1-1955 with nuclear interpretations. Isolated piping sections considered susceptible to thermal overpressurization are analyzed in accordance with the criteria of the ASME Boiler & Pressure Vessel Code, Section NMP Unit 1 UFSAR Section X X-2 Rev. 25, October 2017 III, Appendix F, 1986 Edition. The segment of the system from the recirculation loops to the isolation valves is designed for 1200 psig and 575°F. All equipment in the shutdown cooling system is designed to withstand earthquake acceleration factors of 0.20g horizontal and 0.10g vertical. The heat ex changer tubes, piping, valves and fittings are constructed of stainless steel. The pump casing is made of carbon steel. All components of the shutdown cooling system are located in a common concrete-shielded compartment designed to attenuate radiation levels to 5 mr/hr outside the compartment. 3.0 System Evaluation The cooling water low-pressure section is provided with a number of relief valves and instrumentation, shown on Figure X-1, to protect against overpressurization. The inboard and outboard suction isolation valves are interlocked so that only one valve can be exercised at a time when reactor pressure is above 120 psig. Interlocks also prevent the pumps from operating unless the suction temperature is below 350°F and system isolation valves (38-01, 38-02, and 38-13) are open. During normal reactor operation, the motor-operated isolation valves (38-01, 38-02 and 38-13) are closed and power is removed (to ensure that spurious or inadvertent opening of one of these valves will not result in the loss of the water seal) except when the shutdown cooling system is required to be placed in service. The shutdown cooling system is isolated automatically on a reactor low-low water level signal from the reactor protection system (RPS), described in Section VIII-A. Area temperature detectors are installed at appropriate locations to isolate the system automatically and to initiate an alarm in the control room on any line break. System isolation can also be initiated manually from the control room. Samples may be taken and analyzed for tube leaks from a sample point on the outlet side of the cooling water from the heat exchanger. NMP Unit 1 UFSAR Section X X-3 Rev. 25, October 2017 4.0 Tests and Inspections Once each operating cycle, a physical inspection of the system is made for water leakage with the system in operation. B. REACTOR CLEANUP SYSTEM 1.0 Design Bases The purpose of the reactor cleanup system is to maintain high reactor water purity in order to: 1. Minimize deposition on fuel surfaces by reducing the amount of waterborne impurities in the primary system. 2. Reduce the secondary sources of beta and gamma radiation resulting from the deposition of corrosion products, fission products and impurities in the primary system. 2.0 System Design The cleanup system, shown on Figure X-2, continuously purifies a portion of the recirculation flow and reactor bottom head flow with a minimum of heat loss from the cycle. It can be operated during startup, shutdown, and refueling modes, as well as during normal power operation. Water is normally removed at reactor pressure from one of the reactor recirculation loops and from the reactor bottom head drain line, and cooled in regenerative and nonregenerative heat exchangers, reduced in pressure, filtered, demineralized, and pumped through the shellside of the regenerative heat exchanger to the reactor through the feedwater system. Whenever reactor pressure is insufficient to maintain suction pressure at the main cleanup pumps, an auxiliary pump is used. Two full-size filters and demineralizers of 380,000 lb/hr capacity each are provided to permit continuous operation. Normally, one of these units is in a standby condition. The cleanup filters are pressure-precoat type, and the demineralizers are mixed-bed type. Spent cleanup resins may not be regenerated because of the radioactivity of the impurities removed from the reactor coolant, but may be sluiced from the NMP Unit 1 UFSAR Section X X-4 Rev. 25, October 2017 demineralizer vessels directly to the spent resin tank in the waste disposal system for processing, storing, and eventually offsite disposal. Y-type post-strainers on the inlet and outlet of each of the two demineralizers prevent resins from entering the reactor system in the event of a resin support failure. The regenerative heat exchanger transfers heat from the water leaving the reactor to the water which returns to the reactor. The nonregenerative heat exchanger cools the water further to a normal temperature of 120°F by transferring heat to the RBCLC system at a design rate of 40,000,000 Btu/hr. The nonregenerative heat exchanger is capable of maintaining this low temperature, and not exceeding a maximum of 140°F during blowdown of a portion of the cleanup flow, when effectiveness of the regenerative heat exchanger is reduced. Blowdown is normally used during reactor operation only to remove excess water from the reactor. This blowdown provision is also used as an alternate route for removing refueling water back to the condensate storage tanks (CST) via the main condenser and condensate demineralizers (CND). A cleanup surge tank is provided to assure continuous submergence for the cleanup pumps and to provide a path for the pump recirculation flow. The equipment in this system is designed and constructed in accordance with ASME Pressure Vessel Code, Section III-1965, Class C, ASME Section III-1963, Class C, or ASME Section VIII-1965 with Code Case 1270N. Piping, valves other than those for isolation, and fittings are built to the requirements of ASA B-31.1-1955 with nuclear interpretations. Isolated piping sections considered susceptible to thermal overpressurization are analyzed in accordance with the criteria of the ASME Boiler & Pressure Vessel Code, Section III, Appendix F, 1986 Edition. All equipment in the cleanup system is designed to withstand earthquake acceleration factors of 0.20g horizontal and 0.10g vertical. The equipment and the piping subject to reactor water are constructed of stainless steel, rubber-lined carbon steel, or carbon steel. The primary side of the system subject to the reactor pressure is designed to withstand a pressure of 1300 psig and temperature of 575°F. NMP Unit 1 UFSAR Section X X-5 Rev. 25, October 2017 Operation of the cleanup system is controlled from the main control room. Conductivity of the water entering the cleanup filters and leaving the cleanup demineralizers is recorded in the control room. Filter backwash and resin sluicing operations are controlled from a local panel. Filters are automatically backwashed but are precoated by remote manual operation. All equipment is shielded with concrete except for the main cleanup pumps, surge tank, flow control valve, filter aid tanks and pumps, and precoat tank. The bases for shielding design were determined by the estimated frequency of operating, inspecting and maintaining the various equipment and devices. The shielding is designed to reduce the radiation levels in the valve corridors to 30 mr/hr, and the levels in the access corridors around the cleanup system complex to 5 mr/hr. The unshielded equipment is located in controlled access areas. The hydrogen water chemistry (HWC) and noble metal chemical addition (NMCA or NobleChem) monitoring systems require a continuous supply of reactor coolant at rated reactor pressure and temperature conditions provided from the reactor cleanup system. The monitoring systems include two electrochemical corrosion potential (ECP) monitoring locations, material coupons (durability monitor) for monitoring the noble chemistry coating thickness, the capability for crack growth rate monitoring, and a sample supply for the NMCA temporary equipment needed for the injection of the NMCA material. The ECP monitoring includes a reactor vessel location at a decontamination flange on recirculation loop #12, and the reactor cleanup system location. The ECP/durability/crack growth monitoring and the sample supply is located within the spent fuel pool heat exchanger room in reactor building el 281 ft. The data acquisition system (DAS) for the monitoring of these systems is located on el 281 ft outside the spent fuel pool heat exchanger room. Reactor coolant supply to the monitoring systems is supplied from small bore piping (3/4 in) tie-in connections, installed from the reactor water cleanup (RWCU) 6-in line downstream of the outboard RWCU isolation valve and upstream of the RWCU regenerative heat exchanger. The 3/4-in piping terminates in the spent fuel pool heat exchanger room and a transition is made NMP Unit 1 UFSAR Section X X-6 Rev. 25, October 2017 to 3/4-in stainless steel tubing for supply to the ECP/durability/crack growth monitoring stations. The flow is returned to the RWCU through the reactor water hydrogen chemistry equipment cooler (110-164), and a return flow path through primary containment valve (PCV) 110-163 and blocking valve (BV) 110-167. Figure X-2 shows the system connections to the cleanup system. These monitoring systems require a continuous supply of reactor coolant at rated reactor pressure and temperature conditions provided from the reactor cleanup system. 3.0 System Evaluation The dual full-capacity filters and demineralizers permit maintenance of these units during full-flow operation of the system through the spare units. The filter precoating and filter aid pumps are also duplicated for reliability of the system. The system is provided with relief valves and instrumentation, shown on Figure X-2, to protect against overpressurization of the equipment and overheating of the resins. The cleanup system is isolated by automatically closing the isolation valves on a reactor low-low water level signal from the RPS, described in Section VIII-A. The cleanup system is also automatically shut down and isolated on a system low flow, high pressure or high temperature signal and by area temperature detectors, installed at appropriate locations, which detect line breaks. An interlocking system isolates the cleanup system upon initiation of the liquid poison system. Sample points are provided before and after the cleanup filter and after the cleanup demineralizer to assure the satisfactory performance of equipment. The filter influent sample point is also the source of samples of reactor water during normal operation. A sample point on the cooling water side of the nonregenerative heat exchanger permits analysis for tube leaks. 4.0 Tests and Inspections Isolation valves will be tested periodically to verify operability and leak-tightness, as described in Section VI-D. C. CONTROL ROD DRIVE HYDRAULIC SYSTEM NMP Unit 1 UFSAR Section X X-7 Rev. 25, October 2017 1.0 Design Bases The control rod drive (CRD) hydraulic system is designed to achieve the following objectives: 1. Provide a water source at a constant pressure of about 1,500 psig for charging the scram accumulators. 2. Provide a water source at a constant pressure of approximately 260 psi above reactor pressure to supply water for normal drive operation. 3. Provide a water source at a constant pressure above reactor pressure adequate to supply cooling water for each CRD mechanism. 2.0 System Design The CRD hydraulic system is used to change the position of the neutron-absorbing control rods within the reactor core in response to manual control signals. The system is also designed to scram the reactor in response to manual or automatic signals. The CRD hydraulic system also provides high-pressure makeup to the reactor vessel for a specified leakage of 25 gpm (Technical Specifications) and to provide core cooling in the case of a small line break (up to 0.003 ft2). Each pump can supply 50 gpm water makeup to the reactor vessel. The system minimum delivery rate of 50 gpm within 60 sec will assure that automatic pressure blowdown is not actuated for the specified leakage rate of 25 gpm. The CRD hydraulic system also provides water to the reactor vessel level instrumentation reference leg backfill system (X-C.4.0) and to the keep-full system for the emergency cooling system (Section V-E.2.0). The system, shown on Figure X-3, is made up of supply pumps (two), filters, strainers, control valves and associated instrumentation and controllers. The system is supplied with water from the condensate system (downstream of the CNDs) with two sources of backup. The primary backup source is the CSTs through a check valve. The secondary backup source is the demineralized water storage tank. The water is pressurized, NMP Unit 1 UFSAR Section X X-8 Rev. 25, October 2017 filtered and passed through three pressure reduction stages which regulate output pressures/flow rates. The first pressure stage supplies the highest pressure, nominally 1,500 psig, to the scram subsystem for charging the scram accumulators. The second- and third-stage pressures vary directly with reactor vessel pressure changes. This is accomplished by using valves which develop constant pressure drops due to the constant flow from the first stage. The second and third pressure stages are connected in series and the outlet of the third stage exhausts to the reactor vessel. Therefore, both these pressure stages will vary with reactor pressure changes. The second pressure stage is adjusted to hold a differential pressure, nominally 250-270 psi above reactor pressure, and supplies water for normal drive operation. The third stage supplying cooling water to the drive mechanism is adjusted to maintain approximately 28-44 gpm in the cooling water header. These three supply pressures plus reactor vessel pressure in the exhaust water header are the four operating pressures of the CRD hydraulic system. 2.1 Pumps Two full-capacity centrifugal pumps (one spare) pressurize the system. Normal operation is with one pump running. Switching from one pump to the other is a manual operation. One pump is rated at 85 gpm at a head of 3,760 ft, with a 250-hp motor. The other is rated at 87 gpm at a head of 3,740 ft, with a 250-hp motor. Each pump is installed with a suction strainer and appropriate isolation valves to permit pump maintenance. A minimum flow bypass connection between the discharge of the pump and the CSTs prevents the pump from overheating if the pump discharge valve is inadvertently closed. The pump discharge pressure is indicated at the pump by a pressure instrument. Electric power for this system is normally available from the reserve transformer. Automatic initiation is provided to start each pump from its respective diesel generator in case offsite power is lost. 2.2 Filters The two parallel filters remove 99 percent of foreign material larger than 40 microns from the hydraulic system water. Either NMP Unit 1 UFSAR Section X X-9 Rev. 25, October 2017 filter can be drained, cleaned, and vented for reuse while the other is in service. A differential pressure indicator and alarm monitor the filter element as it collects foreign material. Strainers in the filter discharge lines guard the hydraulic system in the event of a filter element failure. 2.3 First Pressure Stage The first-stage pressure is maintained automatically by a flow-sensing controller (44-145, 44-146, 44-146B) and by an air-operated flow control valve (44-149 and 44-154). By throttling 44-149 or 44-151 to maintain constant flow through 44-145, the pump is caused to operate at the point on its characteristic curve which corresponds to the required pressure. A parallel spare valve is provided with isolation valves to permit maintenance of the noncontrolling valve. During cold shutdown and normal power operation, the CRD system flow control valves, which are normally in automatic mode, may be operated manually. This first pressure stage supplies water to the accumulator charging header. The pressure in this header is monitored in the main control room with a pressure indicator and low and high pressure alarms. 2.4 Second Pressure Stage The second-stage pressure is automatically maintained at approximately 250-270 psi above the reactor vessel pressure by the combined operation of 44-04, 44-179 and 44-178. Valve 44-04 is manually adjusted from the control room so that the flow through it to the reactor produces a drop of approximately 240 psi. When this valve is adjusted, both 44-179 and 44-178 are open. The flow through these valves bypasses 44-04. The bypass flow is adjusted, using flow meter 44-187, to correspond to the flow required by a drive when moving the flow through 44-179 (corresponding to the flow while inserting), and the flow through 44-178 (corresponding to the flow while withdrawing). Electrically, 44-179 is connected so that it closes when the "insert" valves for any drive are actuated, while valve 44-178 closes when the "withdraw" valves for any drive are actuated. In this manner, the flow through these valves always balances the flow to the drives through the 1-in drive water header, the flow through 44-04 is substantially constant, and the required pressure is maintained in the 1-in drive header. The variation in flow requirements between drives is small enough that the NMP Unit 1 UFSAR Section X X-10 Rev. 25, October 2017 corresponding pressure variation is within acceptable limits. Second-stage pressure may be lowered less than 250 psid to compensate for CRD seal leakage or hydraulic control unit (HCU) valve leakage. A standby pair of bypass valves (44-181 and 44-182) are provided along with a manual backup pressure control valve (44-157) used during maintenance on the normal pressure control valves. Filters are installed before the bypass valves to prevent fouling of the bypass valves. Isolation valves are provided for maintenance on the bypass valves. A flow element and an indicator (44-07 and 44-187) are installed for measuring the flow through the bypass valves so that the valves can be adjusted to provide the required flow for normal drive operation. The flow element and the indicator (44-158 and 44-191A) located in the drive water header are used to measure flow to the drives for adjustment and testing. A differential pressure indicator in the main control room shows the differential pressure between the reactor vessel and the drive water header. This pressure indicator is used when adjusting the second-stage pressure with the motor-operated pressure control valve. 2.5 Third Pressure Stage The third-stage pressure is automatically maintained at a pressure above reactor vessel pressure to supply a total of approximately 28-44 gpm cooling water to the drives. The pressure drop which maintains the pressure for this stage is developed by a motor-operated pressure control valve. This valve is manually adjusted from the main control room, and is provided with isolation valves and a manual bypass valve for maintenance. The flow through this valve and the second-stage pressure control valve is substantially constant and the valves, therefore, act to maintain a constant differential above reactor pressure. Changes in the setting of these valves are required only to adjust for changes in the cooling requirements of the drive mechanism as the seal characteristics change with time, and for changes in pump flow characteristics. NMP Unit 1 UFSAR Section X X-11 Rev. 25, October 2017 The cooling water is monitored by a flow indicator. A differential pressure indicator indicates the difference between reactor pressure and cooling water pressure. 2.6 Exhaust Header The exhaust header takes water discharged by the drives during operation and by the third-stage pressure controller and conducts this water to the reactor. The piping is sized to maintain a low differential (approximately 5 psi) above reactor pressure in this header. A check valve permits isolating this line from the reactor vessel and automatically prevents reactor water from flowing into this line should the supply pressure fail. A flow element and an indicator permit measuring the exhaust line flow during Station operation. A bypass line from the pump output to a point upstream of this flow meter allows checking of pump flows. 2.7 Accumulator The accumulator on each drive is an independent source of stored energy to scram that drive. The top of the accumulator contains water; the bottom is initially precharged to approximately 600 psi with nitrogen. To assure that it is always capable of producing a scram, the accumulator is continuously monitored for water leakage and for nitrogen pressure. A float-type level switch will actuate an alarm if water leaks past the nitrogen-water barrier and collects in the bottom of the accumulator. A pressure indicator and a pressure switch are connected to the accumulator to monitor nitrogen pressure. During normal operation the accumulator barrier has virtually zero pressure drop across it. If there should be any loss of nitrogen, the barrier will move onto a stop and further loss will cause a decrease in the nitrogen pressure. The accumulator barrier will not move down beyond the stop and, therefore, will not compress the reduced amount of gas back up to pressure. A decrease in nitrogen pressure will actuate the pressure switch and sound an alarm. An isolation valve allows each of the accumulator instruments to be isolated and serviced. A connection on the accumulator provides for precharging and bleeding. NMP Unit 1 UFSAR Section X X-12 Rev. 25, October 2017 The charging line allows isolation of the accumulator for maintenance and prevents backflow from the accumulator to the charging header. It assures that the accumulator will retain its charge even if the supply subsystem fails. 2.8 Scram Pilot Valves During normal operation, each of the two parallel branches of the RPS energize one of the two three-way solenoid scram pilot valves associated with each drive mechanism. During normal operation, these pilot valves are energized and supply instrument air to the operators of both the inlet scram valve and the outlet scram valve, holding both scram valves closed. During a full scram, both of the RPS branches are de-energized and both pilot valves open, venting the scram valves' operators and allowing the scram valves to open. To protect against spurious scrams, the pilot valves are interconnected so that both pilot valves must be de-energized to vent the scram valves' operators. On the other hand, failure of either electric power to both solenoids or instrument air will produce a scram. The pilot valves are selected based on simplicity of design, a minimum of moving parts, fast opening time, and satisfactory statistical operating history on similar units. For added protection, the instrument air header to all the pilot valves has a pair of backup scram pilot valves. Upon a scram signal these three-way solenoid valves close off the air supply and vent the instrument air header. This will scram any drive should either of its scram pilot valves fail to vent. A diverse reactor trip system, alternate rod injection (ARI), has been added to provide an alternate and diverse method of venting the instrument air header. An ARI initiation signal, high reactor pressure, or low-low water level will actuate the ARI system. 2.9 Scram Valves The inlet scram valve is a globe valve which is opened by the force of an internal spring and closes when air pressure is applied on top of the diaphragm operator. The opening force of the spring is approximately 700 lb. Each valve has a position indicator switch which energizes a light in the control room as soon as the valve starts to open. The scram valve is selected NMP Unit 1 UFSAR Section X X-13 Rev. 25, October 2017 based on high operating force, fast opening time (approximately 0.1 sec) and satisfactory operating history on similar units. Both the inlet and outlet scram valves are similar, except that the inlet scram valve is an angular pattern while the outlet scram valve is a globe pattern. The internal spring preload in the outlet scram valve is slightly greater than the inlet scram valve to provide a faster opening characteristic. 2.10 Scram Dump Volume One scram dump volume is used to limit the loss of and contain the reactor vessel water from all the drives during a scram. The scram dump volume has a capacity to accommodate a free volume of 3.34 gal per drive up to an in-leakage of approximately 0.5 gpm per drive(9). For an in-leakage greater than 0.5 gpm per drive, the free volume will fall below 3.34 gal per drive; however, the system function will be maintained up to an in-leakage of 5 gpm per drive. During normal operation, the dump volume is empty, with both sets of drain and vent valves open. These valves operate very much like the scram valves. With a scram signal, the RPS is de-energized and the two dump volume pilot valves vent the dump volume valves' operators, causing them to close. Position indicator switches on the main valves indicate, with lights in the control room, the position of the vent and dump valves. During a scram, the dump volume partly fills with the water from above the drive pistons. While scrammed, the CRD seal leakage continues to flow to the dump volume until the dump volume pressure equals reactor vessel pressure, or until the scram valves are reset. When the scram signal is removed from the RPS, the scram valves may be closed and the dump volume may be drained. A series of level switches connected to the dump volume indicates by means of annunciators when the volume has emptied after a scram. These level switches also guard against the dump volume being filled such that it cannot accommodate the water discharged during the scram. Should the dump volume start to fill with water, an alarm will sound, and if filling continues, the reactor will automatically scram. A control system interlock will not allow the drives to be withdrawn until the dump volume is emptied to below the scram setpoint. NMP Unit 1 UFSAR Section X X-14 Rev. 25, October 2017 A 1,000-gal capacity holding tank is provided to receive discharge from the scram discharge volume (SDV). The tank is provided with overflow (to reactor building floor drain tank) and level monitoring capabilities. The tank is vented directly to the reactor building ventilation system. A test pilot valve allows the dump volume valves to be tested without disturbing the RPS. Closing the dump volume valves allows the outlet scram valve seats to be leak tested. 2.11 Control Rod Drive Cooling System The cooling system is made up of the third-stage pressure control, the cooling water header, and a check valve which admits water to the underside of the drive piston. Although the drive can function without cooling water, the life of the graphitar seals and elastomer O-rings is shortened by exposure to reactor temperatures; therefore, cooling water is provided to protect these members. When a drive is in motion, the pressure under the piston is higher than the cooling water pressure and the check valve is closed. The check valve opens to admit cooling water when the drive is stationary. 2.12 Directional Control and Speed Control Valves Four solenoid-directional control valves are used for switching the drive water header and the exhaust header to the two drive ports. By energizing and opening two valves at a time, the drive water header can be connected under or over the CRD piston while the exhaust header is connected to the opposite side. Two directional control valves, which include speed control valves, are connected so that they always pass the flow to or from the underside of the piston. This is approximately 4 gpm when the drive is moving at the normal speed of 3 +/- 0.6 in/sec. The balance of forces in the drive mechanism is such that the pressure under the piston is approximately 80 psi whenever the drive is either inserting or withdrawing. Proper speed for control rod insertion is obtained when the speed control element of one of the directional control valves is set, so that a flow of 4 gpm through the valve will produce a pressure drop of 170 psi (from 250 psi in the drive-water header to 80 psi under the piston). Similarly, for control rod withdrawal, the other directional control valve is set so that 4 gpm will produce a NMP Unit 1 UFSAR Section X X-15 Rev. 25, October 2017 pressure drop of 75 psi (from 80 psi under the drive piston to 5 psi in the exhaust header). The directional control valves are protected from dirt by filters. (Note: 3 cps rod speed corresponds to 48 +/- 9.6 sec stroke time.) The cooling, control, and scram systems for each drive use common piping to the drive; therefore, the directional control valves are periodically subject to scram pressure. The two directional control valves connected to the drive water header can be caused to open if subjected to this higher pressure on their outlet ports. A check valve prevents significant loss of water to the drive water header during scram. 2.13 Rod Insertion and Withdrawal As described in preceding paragraphs, insert motion is obtained by opening the proper pair of valves. In order to unload the collet so it can be unlocked, this pair of valves is also opened for approximately one-half sec during a withdrawal operation. After this a pair of withdraw valves are opened. This is accomplished by an electrical sequence timer and occurs automatically when the rod withdraw operating switch is closed. "Jogging" is accomplished in a similar manner. When this mode of operation is selected by the Operator, the proper pair of valves is energized electrically long enough to allow the drive to move to the next notch position, at which time the valves are automatically de-energized, even if the Operator holds the switch closed. This feature relieves the Operator of having to estimate the time required to accomplish a single notch movement. If all four directional control valves are closed while the drive is in a position between notches, water displaced by the drive piston must leak past the drive seals in order for the drive to settle into latched position. With normal seals, including those well worn, this settling speed is a fraction of normal withdraw speed. To speed up settling and latching, the settle circuit delays the closing of the withdraw valve for approximately 5 sec. This allows the drive to withdraw at about one-half normal speed to the next latch position. The jog withdraw time interval is shortened so that the settle cycle begins before the drive has withdrawn a full notch. NMP Unit 1 UFSAR Section X X-16 Rev. 25, October 2017 2.14 Scram Actuation During a scram, a separate set of valves are utilized. These valves (inlet and exhaust scram valves) open, admitting the pressure in the accumulator (approximately 1,500 psi) under the main drive piston, and venting the area over this piston to the scram dump volume. This volume is maintained at atmospheric pressure during normal operation. The large differential pressure (initially about 1,400 psi and always several hundred psi--depending on reactor pressure) produces a large upward force on the index tube and control rod, giving the rod a high initial acceleration and providing a large margin of force to overcome any possible friction or binding. This initial scram force is a maximum of 6,000 lb under cold reactor conditions and 2,800 lb when the reactor is at operating pressure. The characteristics of the hydraulic system are such that after initial acceleration (less than 30 msec after start of motion), the desired scram velocity of about 5 ft/sec is achieved and the drive travels at a fairly constant velocity. This characteristic provides a high initial rod insertion rate and a high operating force margin that cannot be achieved by a drive designed to utilize gravity forces. As the drive piston nears the top of its stroke, the piston seals close off the large passage in the exhaust line and the drive velocity is reduced. Each drive requires about 2.5 gal of water during the scram stroke. There is adequate water capacity in each drive's accumulator to complete a scram in the required time at low reactor pressures. At higher reactor pressures, the accumulator is assisted by reactor pressure reaching the drive through a ball check valve located in the drive itself. As water is drawn from the accumulator, the accumulator discharge pressure falls below reactor pressure. This causes the check valve to shift its position to admit reactor pressure under the drive piston. Thus, reactor pressure furnishes the force needed to complete the scram stroke at higher reactor pressures, while the accumulator alone will accommodate the low-pressure scrams. When the reactor is up to full operating pressure, the accumulator is not required to meet scram time requirements. With the reactor at 1,000 psig, the scram force is still over 1,000 lb without an accumulator. 3.0 System Evaluation NMP Unit 1 UFSAR Section X X-17 Rev. 25, October 2017 3.1 Normal Withdrawal Speed Normal withdrawal speed is determined by differential pressures at the drive, and is set for a nominal value of 3 in/sec. The characteristics of the pressure regulating system are such that this speed is maintained independent of reactor pressure. Tests have determined that accidental opening of the speed control valve to the full open position will produce a velocity of approximately 6 in/sec. Should this system fail, producing maximum available pump pressure (1,750 psig) to the drive system with zero reactor pressure, the hydraulic resistances in the system would limit the withdraw velocity to 2 ft/sec. The allowable operating limits on withdraw and insert speed are determined by requirements for the insert-before-withdraw motion and for jogging. These limits are lower than those which might be set by considerations of maximum allowable reactivity variations. The jog withdraw operation of the drive is an excellent test of the correctness of the speed setting; the drive generally will fail to withdraw if the speed is incorrectly adjusted. A pressure of approximately 60 psi higher than reactor pressure must be maintained above the main drive piston in order to keep the collet unlocked. This corresponds to a pressure greater than 60 psi above reactor pressure under the main piston. Any malfunction which allows the pressure to drop below this value, a condition necessary for higher withdraw speeds, results in collet locking. 3.2 Accidental Multiple Operation Each drive mechanism has its own complete set of electrically operated directional control valves, which are closed when de-energized. The correct operation of all four valves in the correct sequence is required to cause the drive to withdraw. Consequently, the probability of multiple, simultaneous, independent valve failures that could cause accidental multiple rod withdrawal is extremely small. The electrical system which actuates the directional control valves is designed to prevent any credible failure from producing accidental movement of more than one control rod. 3.3 Scram Reliability NMP Unit 1 UFSAR Section X X-18 Rev. 25, October 2017 High scram reliability is the object of a number of features in the system, such as the following: 1. There are two sources of scram energy (accumulator and reactor pressure) for each drive whenever the reactor is operating. 2. Each drive mechanism has its own scram valves and pilots so that only one drive can be affected by a scram valve failure to open. A separate backup pilot valve is provided to scram a drive (after some time delay) should this failure occur. 3. Under scram conditions, the drive mechanism develops from 6,000 lb (at zero reactor pressure) to 2,800 lb (at rated pressure) of force, a large margin to overcome possible friction. 4. The scram system and mechanism are designed so that the scram signal and mode of operation override all others. 5. The scram valves fail open on loss of either air or electrical power. Hence, failure of the valves, air system or electric system will generally produce, rather than prevent, a scram. All components used in the scram hydraulic system are selected either after an extensive testing program or after many millions of accumulated operating hours in service. 3.4 Operational Reliability High operational reliability contributes generally to overall safety by minimizing the occasions when abnormal operating conditions are encountered. High operational reliability is the objective of the following features of the CRD hydraulic system. 1. Components in the hydraulic system are picked based on established reliability. A spare pump and control valves are provided for reliability. Operating valves are accessible for maintenance while the reactor is in operation. NMP Unit 1 UFSAR Section X X-19 Rev. 25, October 2017 2. Provisions are made to operate with a reasonable amount of foreign material in the reactor water and in the water supplied to the hydraulic system. Filters and strainers are incorporated in the drive mechanism in passages through which water is drawn into the mechanism. Filters, hydraulic system downstream of the supply pump, and the positions in the hydraulic system protect the operating valves. 3.5 Alternate Rod Injection The ARI system provides a redundant (to the RPS) and diverse means of control rod insertion. This system uses sensors (reactor high pressure and reactor low-low water level) that are independent of the sensors that initiate reactor scram. The ARI valves are dc operated and energized to vent to provide diversity from the scram pilot valves. 4.0 Reactor Vessel Level Instrumentation Reference Leg Backfill The CRD system supplies water to the reactor vessel level instrumentation reference leg backfill system. This system injects a small flow into three level instrument reference legs to prevent the accumulation of noncondensible gases. The gases can cause erroneous measurements on reduction in reactor pressure. The supply is taken from the CRD water pumps discharge header using the former reactor head spray piping. The head spray flow control valve is removed, and the backfill supply is taken from the downstream side of the flow element. The reactor head spray cooling system has been removed from service by permanently removing the removable spool piece. All head spray piping inside the drywell has been removed and capped off at drywell penetration. Appendix J Type A testing of the penetration assures primary containment is maintained. 5.0 Tests and Inspections The pumps and flow control valves are operated alternately with their spare units, providing observation of performance. NMP Unit 1 UFSAR Section X X-20 Rev. 25, October 2017 Instrumentation and alarms monitor operation of flow and pressure regulation to assure availability of drive water and cooling water. Operation of drive control, scram and scram pilot valves is observed during periodic testing of CRD operation and during scram tests. CRD hydraulic pumps are tested in accordance with the Nine Mile Point Nuclear Station - Unit 1 (Unit 1) Technical Specifications. The ARI and backup scram valves are tested during each refueling outage. Either a postscram walkdown of the SDV header and instrument volume or a hydrostatic test and visual examination, in accordance with ASME Section XI, is performed once per refueling cycle to satisfy the recommendations in Generic Letter 86-01 regarding scram system piping integrity. D. REACTOR BUILDING CLOSED LOOP COOLING WATER SYSTEM 1.0 Design Bases The RBCLCW system provides demineralized water at temperatures not exceeding 95°F to cool auxiliary equipment located in the reactor, turbine and waste disposal buildings. The closed loop permits isolation of systems containing radioactive liquids from the service water, which is returned to the lake. The cooling load imposed on the system will largely depend on the Station power output at any given time; therefore, this system has sufficient capacity and flexibility to cool various combinations of equipment regardless of the Station power output. With a RBCLCW flow of 8500 gpm, a total service water flow of 10,000 gpm, and two RBCLC heat exchangers in service, the RBCLC system was designed (i.e., heat exchangers sized, built and procured) such that the system would have a nominal heat removal capability of approximately 126 x 106 Btu/hr. This value is not the RBCLC system cooling capacity requirement. The above performance is dependent on temperatures (RBCLCW temperature and the temperature of the ultimate heat sink (UHS) - Lake Ontario) as well as tube and shell flows. This value does not represent the cooling capacity necessary for plant operation. The cooling load imposed on the system will largely NMP Unit 1 UFSAR Section X X-21 Rev. 25, October 2017 depend on the Station power output and plant condition. The RBCLC heat removal requirements for the most limiting modes of operation are as follows: Modes of Operation Heat Load [106 Btu/hr] Normal Operation 74.24 Normal Shutdown 97.82 10-hr Shutdown 156.20 2.0 System Design The RBCLCW system provides cooling water to the following major components (Figure X-4). Fuel Pool Heat Exchangers Instrument Air Compressors Electric Feedwater Pumps Condensate Pumps Feedwater Booster Pumps Control Room and Laboratory Air Conditioning Equipment Recirculation Pump Coolers Cleanup System Nonregenerative Heat Exchangers Reactor Building Equipment Drain Tank Cooler Drywell Air Coolers Waste Disposal System Heat Exchangers Shutdown Cooling System Heat Exchangers and Pump Coolers Offgas Vacuum Pump Coolers The system consists of three horizontal centrifugal pumps rated at 4500 gpm with a total developed head (TDH) of 65 psi each, and three counterflow shell and tube heat exchangers, plus NMP Unit 1 UFSAR Section X X-22 Rev. 25, October 2017 necessary flow control valves, instrumentation and piping. During normal Station operation, one or two pumps may be operated depending on the system heat loads (cooling requirements) and lake temperature. For the most demanding load cases, i.e., 10-hr and normal shutdown, any combination of one RBCLC pump and two RBCLC heat exchangers, or two RBCLC pumps and three RBCLC heat exchangers, will provide adequate cooling, i.e., RBCLC effluent temperature of 90 +/- 5°F and sufficient flow to required on-line users. These combinations are intended to limit RBCLC heat exchanger shellside flow to approximately 3000 gpm per heat exchanger to prevent flow-induced tube vibration. Temperature indication of each component and flow indication on major lines help to maintain the proper amount of cooling water to each component. These indications are either local or in the main control room. The service water for the RBCLC system (tubeside of heat exchangers) is supplied by the service water (SWP) system utilizing two normal service water pumps and backed up by two emergency service water pumps. Additional low-conductivity water can be added to the system from the 2000-gal closed loop cooling makeup tank (Figure X-5) located on floor el 351' in the turbine building. The closed loop cooling makeup tank is shared with the turbine building closed loop cooling (TBCLC) system and provides a low-pressure inlet of low-conductivity water to both systems. The tank is supplied with water automatically from the condensate transfer system through a makeup level control valve. Additional makeup is also available from the makeup demineralizer tank. Excess water due to thermal expansion in the RBCLC system will overflow through an elevated drain into the turbine building equipment drain sump. To facilitate maintenance activities, the RBCLC system is designed for flexibility of operation. Each of the three RBCLC system pumps and heat exchangers may be interchanged as necessary. Each RBCLC pump is normally started and stopped from the control room. For normal operation, one or two pumps supply the cooling water requirements. NMP Unit 1 UFSAR Section X X-23 Rev. 25, October 2017 The cooling water pumps and the heat exchangers are designed to withstand seismic forces of 0.26g horizontal and 0.13g vertical. Heat exchangers are constructed in accordance with the ASME Boiler and Pressure Vessel Code, Section VIII-1986. Pumps conform to the ASME Power Test Code for Centrifugal Pumps. All equipment piping connections are ASA standard. System piping is designed in accordance with Sections 1 and 6 of the ASA B31.1-1955 Code. 3.0 Design Evaluation The most demanding heat load case on the RBCLC system is the 10-hr shutdown. Assuming the most conservative RBCLC system lineup, i.e., one RBCLC pump, two RBCLC heat exchangers, total service water flow of 9000 gpm to RBCLC system, and lake water temperature equal to 83°F, the RBCLC system will be able to reject enough heat to maintain a mixed mean temperature below 95°F. RBCLC heat exchanger shellside flow is limited to 3000 gpm to prevent flow-induced tube vibration. Service water velocity within the tubes is normally maintained at not less than 4 fps to minimize tube fouling due to sand and silt (except when operating in the emergency shutdown mode). The emergency shutdown mode is an off-normal condition that addresses the possibility of failure of the controller to the RBCLC heat exchanger temperature control valve. Only the RBCLC essential heat loads associated with accident mitigation were considered. Other design assumptions for the assessment include single-pump, two-heat-exchanger operation. Under this scenario, the tubeside velocities will be less than 4 fps. Since this scenario is a design basis accident (DBA) event that occurs during an off-normal emergency shutdown, it is expected that long-term cooling would be provided for at least 30 days, which, due to the limited duration and conservative analysis assumptions, is unlikely to be impacted by low tube velocities. If the quantity of service water at operating velocities should tend to chill the cooling water below approximately 85°F, a bypass piping arrangement with flow control valves will divert some RBCLC water around the heat exchangers, remixing it downstream to maintain the set temperature. Two temperature-controlled flow control valves regulate the volume of cooling water entering the shellside of the heat exchangers. Operating in tandem, one valve will admit cooling water to the NMP Unit 1 UFSAR Section X X-24 Rev. 25, October 2017 heat exchanger supply manifold and the other will divert the cooling water to the discharge header. As flow to the supply header is diminished, the diverted water flow is increased. A mechanical travel stop in the actuator of the supply header cooling water temperature control valve limits the valve from closing completely to assure heat removal capability during a DBA loss-of-coolant accident (LOCA) event, coincident with a loss of control signal or air. Although not credited for safety function, a trip valve in the air line to the actuator will lock the valve in its last position, thereby providing time to establish manual control of the valve. A temperature element in the cooling water discharge manifold from the heat exchangers actuates the SWP and cooling water control valves. The SWP and RBCLC flow control valves may be manually bypassed/operated in order to maximize cooling and control flow. Manual override of these valves would only be necessary when shellside heat exchanger flow needs to be limited while tubeside heat exchanger flow (service water) needs to be increased. To evaluate leakage from equipment into the closed loop, the outlet of each major component on the cooling water system is provided with a grab sampling station. Leakage out of the system is noted by a flow switch and flow alarm in the system makeup line. Major components served by the cooling water system are provided with high temperature alarms and/or temperature transmitters to aid in regulating cooling water flow. In the event of the loss of normal and reserve ac power, two of the RBCLC pumps are connected to power board (PB) 16 and one to PB 17. These power boards are supplied power from diesel generators in the event of failure of their normal supply, as described in Section IX, Electrical Systems. The emergency service water pumps are also powered by the diesel generators in order to maintain a supply of cooling water to the RBCLC heat exchangers. 4.0 Tests and Inspections The standby pump(s) are operated periodically to assure that they function properly. NMP Unit 1 UFSAR Section X X-25 Rev. 25, October 2017 Drywell isolation valves on the cooling water system are exercised periodically to assure proper operation. E. TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM 1.0 Design Bases The function of the turbine building closed loop cooling water (TBCLCW) system is to provide demineralized cooling water in a closed loop to auxiliary equipment in the turbine building. The closed loop permits isolation of systems containing radioactive liquids from the service water which returns to the lake. The cooling water temperature is maintained at temperatures not exceeding 95°F. The cooling load imposed on the system will largely depend upon the Station power output at any given time. Therefore, this system has the flexibility to accommodate a wide range of variations. The system has a heat removal capability 52.34 x 106 Btu/hr and a flow capacity of 11,000 gpm. (Although the original design parameters were used in sizing of replacement heat exchangers, the original heat removal capability of 52.5 x 106 Btu/hr was reduced to 52.34 x 106 Btu/hr by the application of more conservative fouling factors.) This capability is based upon using two of three heat exchangers (heat exchangers nos. 11 and 12). Heat exchanger number 13 has a 10 percent higher heat removal capability than the nos. 11 and 12 heat exchangers. 2.0 System Design The system consists of two full-capacity centrifugal pumps (each rated at 11,000 gpm at a TDH of 61 psi), and three half-capacity heat exchangers plus necessary flow control valves (Figure X-5). Cooling water is supplied to the following equipment: Shaft-Driven Reactor Feedwater Pump Oil Tank Coolers Steam Packing Exhauster Coolers Mechanical Vacuum Pump Coolers Hydrogen Coolers NMP Unit 1 UFSAR Section X X-26 Rev. 25, October 2017 Stator Coolers Generator Lead Coolers Recirculating Pump Motor Generator (MG) Set Coolers House Service Air Compressor Sample Coolers Turbine Building Equipment Drain Tank #11 Instrument Air Compressor #13 Instrument Air Dryer #12 Battery Room #14 Air Conditioner Condensate Filtration System Air Compressor Temperature indication of each component and flow indication on major lines helps to maintain the proper flow of coolant to each component. Each turbine building cooling water pump is normally started and stopped from the control room. For normal operation, one pump will supply the cooling water requirements. The second pump will be started manually when required. Service water, when necessary, is supplied by the SWP system (Section X-F) utilizing two service water pumps for the plant. Additional low-conductivity water can be added to the system from the 2000-gal closed loop cooling makeup tank (Figure X-5), located above the pump suction manifold at el 351. The closed loop cooling makeup tank is shared with the RBCLC system and provides a low-pressure inlet for makeup water in both systems. Makeup water to the tank is automatically supplied from the condensate transfer system through a makeup level control valve. Additional makeup water is also available from the makeup demineralizer tank. Excess water due to thermal expansion in the TBCLC system will overflow through an elevated drain into the turbine building equipment drain sump. NMP Unit 1 UFSAR Section X X-27 Rev. 25, October 2017 The TBCLC system is designed for flexibility of operation, which permits the use of any combination of pumps and heat exchangers, and also to facilitate maintenance. The cooling water pumps are designed to withstand the following seismic acceleration forces. Horizontal 0.15g Vertical 0.075g Heat exchangers are constructed in accordance with the ASME Boiler and Pressure Vessel Code, Section VIII-2001 with 2003 addenda (heat exchanger no. 13) and Section VIII-1983 (heat exchanger nos. 11 and 12). Pumps conform to the ASME Power Test Code for Centrifugal Pumps. All equipment piping connections are ASA standard. 3.0 Design Evaluation Either pump and any two heat exchangers have the capacity to cool the entire system with a maximum 77°F lake water temperature.* Service water velocity within the cooling water heat exchanger tubes is maintained at a minimum velocity of 4 fps to minimize fouling due to sand and silt. In the event that the quantity of service water at this velocity should tend to chill the cooling water, a bypass piping arrangement with flow control valves diverts some TBCLCW around the heat exchangers, remixing it downstream to maintain the set temperature less than or equal to 95°F. Two temperature-controlled flow control valves regulate the volume of TBCLCW entering the heat exchangers. Operating in tandem, one valve will admit TBCLCW to the cooling water heat exchanger supply manifold, and the other will divert the TBCLCW to the discharge header. As flow to the supply manifold is diminished, the diverted water flow is increased. A temperature element in the TBCLCW discharge manifold from the heat exchangers actuates the SWP and TBCLCW control valves. The SWP control valves located in the SWP discharge manifold can be bypassed by manual control.

• This is a design point reflecting system capacity at 77°F.

NMP Unit 1 UFSAR Section X X-28 Rev. 25, October 2017 To evaluate radiation hazards as a result of leakage from equipment into the cooling water system, the outlet of each major component on this system is provided with a grab sampling station. High temperature alarms and temperature transmitters for major components served by the cooling water system aid in regulating cooling water flow. Excessive leakage out of the system is noted by a flow switch and alarm in the system makeup line. 4.0 Tests and Inspections The alternate cooling water pump is exercised periodically to assure its proper operation. F. SERVICE WATER SYSTEM 1.0 Design Bases The purpose of the SWP system is to provide strained lake water for cooling the RBCLCW and TBCLCW systems, the steam jet air ejector (SJAE) precooler, ejector vent cooler, the building local air coolers and other building services. Service water also is supplied to the screenwash pumps, the radwaste solidification and storage building (RSSB), and the makeup demineralizer. The system is to be available to cool the reactor building cooling water system under all conditions of operation. The cooling water requirement during the shutdown mode represents the most severe condition and is used as the design basis. 2.0 System Design The system is shown on Figure X-6. Lake water from the intake tunnel passes through trash racks and traveling screens in the screen and pump house and floods the service water pump well. Two full-capacity (20,000-gpm) vertical sleeve bearing pumps take suction from the well. Each pump is provided with a .03-in mesh automatic self-cleaning strainer. The pump discharges are passed through the self-cleaning strainers and blocking valves and then into two separate headers which deliver water to cooling loads within the plant. A valved crosstie located NMP Unit 1 UFSAR Section X X-29 Rev. 25, October 2017 downstream of the pumps enables either pump to supply either service water header. In the reactor building, the SWP system provides flow to the RBCLC heat exchangers. Downstream of the RBCLC heat exchangers is temperature control valve (TCV) 72-146. This valve regulates the amount of service water flowing through the heat exchangers. The RBCLC system provides the control signal for the valve's position. In parallel with the TCV is bypass valve 72-92R, for use if the TCV is out of service or to increase flow through the heat exchangers during peak loading conditions, as shown on Figure X-4. Listed below are the systems and requirements fulfilled by the SWP system. RBCLCW Heat Exchangers (Tube Side) 6,200 gpm TBCLCW Heat Exchangers (Tube Side) 8,000 gpm Screenwash System Pumps 2,400 gpm SJAE Precoolers and Vent Cooler 1,000 gpm Local Building Area Coolers 1,500 gpm RSSB HVAC Chillers 650 gpm Makeup Demineralizer 100 gpm Breathing Air Compressor 7 gpm Total 19,857 gpm To provide for future added capacity, the pump header was extended and two valved branches were added. In the event of a loss of both normal and reserve ac power, the service water pumps would be unavailable. At this time, service water requirements for the RBCLCW heat exchangers would be met by either of a pair of emergency ac power vertical turbine pumps. One of these pumps is connected to PB 16 and the other to PB 17. These power boards are supplied power from the diesel generators if their normal supply fails, as described in Section IX, Electrical Systems. The emergency pumps, each rated at 3,600 gpm, are in the screenhouse and take their suction from the circulating water intake. NMP Unit 1 UFSAR Section X X-30 Rev. 25, October 2017 Each of the emergency service water pumps is connected to one of the service water supply lines to the RBCLCW heat exchangers in the reactor building. Each emergency service water pump can supply water to any one of the three heat exchangers. Each of the TBCLCW heat exchangers is serviced by two full-size service water supply lines. During normal operation, both the supply headers on the RBCLC side and TBCLC side are engaged by keeping the blocking valves open; however, to perform maintenance or other plant activities, one of the RBCLC side and one of the TBCLC side blocking valves can be secured. 3.0 Design Evaluation Either pump has the capacity with the bypass valve opened to provide maximum service water requirements and can be throttled safely to flows as low as 20 percent of design if a need arises to reduce flow for temperature control. With bypass valve 72-92R closed, two normal service water pumps are required to meet the required flow rates for the most limiting mode of system operation. The two emergency service water pumps increase the reliability of the SWP system and, as previously mentioned, provide service water during loss of normal and reserve ac power. In the unlikely event both emergency service water pumps fail to operate (i.e., due to a fire), an intertie exists between the diesel fire pump and the emergency service water line. The diesel fire pump is capable of handling the additional emergency service water requirements. The double supply lines to the closed loop cooling water heat exchangers provide 100-percent backup in the event of pipe failure in either building. A minimum velocity of 4 fps is maintained in both the RBCLCW and TBCLCW heat exchangers tubes to deter sand buildup, except for the RBCLCW when operating in the emergency shutdown mode. In the event of loss of a SWP pump, low service water header pressure will be alarmed in the control room and the alternate pump will be started manually. NMP Unit 1 UFSAR Section X X-31 Rev. 25, October 2017 Differential pressure alarms across all strainers signal excessive pressure drop to the Operator in the control room. IE Bulletin 80-10 requires effluent radiation monitoring for those systems that are normally considered nonradioactive, but could possibly become contaminated by leakage from interfacing systems. The 42-in reactor building service water return and 10-in turbine building service water return lines are alternately monitored for radiation at 15-min intervals prior to discharge. Other service water return lines, including cooling to the RSSB air conditioning units which have no credible potential for contamination, are not monitored for radiation prior to discharge. 4.0 Tests and Inspections To assure its availability, the alternate pump is operated periodically. Both emergency service water pumps are operated quarterly. G. MAKEUP WATER SYSTEM 1.0 Design Bases The makeup demineralizer system is a truck-mounted portable system that is normally parked in the turbine building. It normally receives its supply water from the SWP system. Backup water from the city water system is available. The system was designed to deliver batches of demineralized water to fill the demineralized water makeup tank, the CSTs, and other reservoirs (e.g., the waste surge tank) as necessary. 1. Capacity: a. 100-150 gpm continuous b. Until the CSTs and the demineralized water makeup tank are filled, up to approximately 335,000 gal. 2. Quality: NMP Unit 1 UFSAR Section X X-32 Rev. 25, October 2017 a. Conductivity: < 0.1 Micromhos/cm3 or < .1 uS/cm b. TOC: < 400 ppb c. Silica: < 10 ppb d. Chlorides: < 10 ppb e. Sulfates: < 10 ppb 2.0 System Design The raw water taken from the discharge side of the Station service water pumps passes through a precipitator and clearwell to either of the demineralizer feed pumps. The system processes water at a rate of approximately 150 gpm, so it routinely visits the Station for periods of several days to replenish the demineralizer water storage tank and the CSTs. Since the minimum allowable CST volume is 105,000 gal, the portable makeup system is not required to replenish more than 295,000 gal depleted from the CSTs, and another 40,000 gal for a depleted demineralized makeup water storage tank. The makeup system also has a flanged connection upstream of the retired demineralizers that may be used for connection of a portable skid-mounted (in-plant), as well as the truck-mounted, small-capacity demineralized water unit. The portable skid- or truck-mounted system typically consists of charcoal filters, followed by demineralizer banks (cation, anion, and mixed). The truck-mounted demineralizer is dispatched with a custom loaded resin charge for a specific influent water supply. The exact configuration used may vary depending upon demineralizer influent water chemistry quality. The raw water is taken from the clearwell via either no. 11 or no. 12 demineralizer feed pump. Connection from the pump discharge to the portable demineralizers inlet is effected through a 2-in flexible hose. NMP Unit 1 UFSAR Section X X-33 Rev. 25, October 2017 The demineralizer effluent goes to the makeup demineralized water storage tank. Connection from the portable demineralizer effluent to the storage tank inlet line is made using a 2-in flexible hose. Diversion of the portable demineralizer discharge to the drain during startup is possible. A discharge line sampling point is also provided. Operation of the portable makeup demineralizer is manually initiated at the makeup demineralizer system control panel, the skid-mounted or truck-mounted unit, and at various places in the makeup system. Demineralized effluent goes to the demineralized water storage tank on el 369, which has a capacity of 40,000 gal, and the CSTs on el 261, which have a combined capacity of 400,000 gal. The demineralized water from these tanks can be used to provide an alternate source for the following: Internals Storage Pit Head Cavity Cleanup System Control Rod Drive System Resin Transfer and Regeneration Equipment Spent Fuel Pool Chemical Addition Tank Radiation Monitor Flush Line Closed Loop Cooling Makeup Tank Main Condenser Demineralized water is normally provided directly to the following: Liquid Poison System Laboratories and Sample Sinks NMP Unit 1 UFSAR Section X X-34 Rev. 25, October 2017 Stator Winding Liquid Cooling System Condensate Filtration System Air Compressor 3.0 System Evaluation Operation of the portable makeup system is on demand at routine infrequent intervals to replenish demineralized water in storage tanks. With the system inoperable or when the portable demineralizer skid is not available, the Station can continue operation with makeup water from the CSTs which have a combined capacity of 400,000 gal. Additional makeup water is available from the demineralized makeup water storage tank which has a 40,000-gal capacity. As an option, Operators may take a supply of water from city water for processing, depending on the plant operating conditions. City water is an equivalent or better source for makeup than lake water in terms of contaminants, and delivery capacity is within or exceeds the requirements for supply to the demineralized water system. 4.0 Tests and Inspections The demineralizer effluent is controlled by effluent conductivity, but periodic samples are taken of conductivity, TOC, silica, chlorides, and sulfates. H. SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM 1.0 Design Bases This system is designed to remove the spent fuel assemblies' decay heat and the impurities from the pool water so as to maintain the temperature and purity of the spent fuel pool water at acceptable levels, assuring clarity under all anticipated conditions. The pool water temperature is maintained at or below 140°F during maximum anticipated storage conditions and 110°F during reactor power operation to maintain the secondary containment licensing basis. Normal refueling conditions are based on refueling the reactor every 24 months. During certain NMP Unit 1 UFSAR Section X X-35 Rev. 25, October 2017 instances, it may be necessary to offload the entire core into the spent fuel pool. The maximum heat generation rate was determined by assuming a full core discharge (532 bundles) after 24 months, with the maximum number of previously discharged fuel bundles (3550) being present in the pool. The greatest portion of the decay heat would be produced by the bundles being discharged from the core, rather than those bundles which have been stored in the spent fuel pool from previous discharges. The long-term decay heat rate for GE11 and GNF2 fuel is essentially the same as for previous fuel designs. Therefore, the decay heat rate used as the basis for the spent fuel storage pool filtering and cooling system design remains unchanged. Prior to Technical Specification Amendment No. 167, the spent fuel pool was licensed for 2776 storage cells. The north half of the pool contained 1066 nonpoison flux trap storage locations, and the south half provided 1710 locations using Boraflex as a neutron absorber. Currently, the spent fuel pool is licensed, per Technical Specification Amendment No. 167, for 4086 spent fuel storage locations using the neutron absorber material Boral, with 1840 storage locations in the north half of the pool and 2246 locations in the south half. The nonpoison racks in the north half of the pool were replaced with new poisoned racks after the 1999 refuel outage. The reracking of the south half of the pool has been partially completed. Six of the eight existing Boraflex racks have been replaced with new Boral racks, increasing the capacity from 1296 to 1656 storage locations. Two Boraflex racks remain in the south half, providing 414 storage locations. The rerack of the remaining two racks has been deferred until further capacity increase is warranted. Unit 1 committed to the Nuclear Regulatory Commission (NRC) that refueling and core offloading operations would not begin until it was determined that the spent fuel pool cooling systems were operable, to ensure that the bulk pool temperature limits would not be exceeded. For a normal (full core offload or core shuffle) refueling, the offload time to the spent fuel pool and the RBCLC temperatures shall be verified to be consistent with a bulk pool temperature not to exceed 140°F with one cooling train operating. NMP Unit 1 UFSAR Section X X-36 Rev. 25, October 2017 For the case of an abnormal maximum heat load (such as a full core offload shortly after a normal refueling), this would require verifying that offload time and RBCLC temperatures were consistent with a pool temperature <140°F with both cooling trains operating. Based on past experience, sufficient clarity of the pool water can be achieved by a filter capable of removing particles as small as 25 microns in size. 2.0 System Design The system is shown on Figure X-8. Two full-capacity (600 gpm) pumps take suction from the pool surge tanks and circulate the pool water through two parallel loops consisting of one filter and one heat exchanger. The water is returned to the pool on the side opposite the surge tank skimmers. The spent fuel pool cooling (SFC) system is designed as seismic Category 1. The SFC system bounding design conditions are that, under full core discharge conditions with RBCLC coolant water temperature at its maximum of 95°F, and assuming the SFC heat exchangers are fouled to their design maximum and 5 percent of the tubes are plugged, a pool water temperature of 140°F would be reached if a full core offload began 1008 hr after reactor shutdown, and was completed 1129 hr after reactor shutdown with one of the two redundant cooling trains operating. A more expedited offload may be performed if the plant conditions exist to maintain the pool water temperature at or below 140°F with one SFC train operating. Flow control valves regulate the flow in each loop at 600 gpm by use of a controller that may be operated in the auto or manual mode. Cooling water is supplied to the heat exchangers from the RBCLCW system at temperatures not exceeding 95°F. A sample point is incorporated to determine any tube leakage. Initial filling and level maintenance in the spent fuel pool and surge tanks was from the condensate transfer system. The total NMP Unit 1 UFSAR Section X X-37 Rev. 25, October 2017 volume of the surge tanks is approximately 2000 cu ft. They will normally run at a level of approximately 1000 cu ft. The difference in surge tank volume allows for the displacement of water from the spent fuel storage pool when a shipping or transfer cask (or any other object) is placed in the pool. Makeup water is provided by the condensate transfer system. Normally, makeup is directly to the spent fuel storage pool. Makeup to the spent fuel storage pool is automatically initiated when the surge tank volume decreases to 800 cu ft and stops when the volume reaches 1000 cu ft. If the makeup to the spent fuel storage pool is not sufficient to maintain surge tank volume, makeup water can be provided directly to the surge tanks. The condensate transfer system can provide a makeup rate of 75 gpm or more to either the spent fuel storage pool or the surge tanks. Makeup water can also be supplied directly to the spent fuel pool through fire water hoses. Any particles that enter the pool either sink to the bottom to be removed by a portable vacuum cleaner or float about in the pool and eventually enter the skimmers, surge tanks and filtering loop. Provision is made for transferring water to the liquid waste disposal system for processing if the pool water becomes highly contaminated. The precoat-type filters use porous carbon elements. Precoat material is powdered/crushed resins. One precoat mix tank and pump serves both filters. The slurry is circulated through the filter vessel and back to the tank until a uniform coating of precoat material covers all the elements. The filter is then placed in service until differential pressure signals the need for backwashing. The backwashing process consists mainly of first valving off and draining the filter, then filling the filter with condensate from the condensate transfer system. All vents are closed during this filling and air is trapped in the filter dome above the elements. When the pressure in the filter dome reaches approximately 80-100 psig, the drain valve is quickly opened and the filter cake, together with trapped impurities, washes into the fuel pool filter sludge tank. From the sludge tank the suspension of impurities and water is pumped to the waste disposal system. Aside from its normal function of cooling and purifying the spent fuel pool water, the system is also used after reactor NMP Unit 1 UFSAR Section X X-38 Rev. 25, October 2017 refueling to drain the reactor internals storage pit and head cavity. Alternate lines allow transport of the water to either the main condenser or to the waste disposal system for processing. In either case the water is filtered, demineralized and returned to the CSTs. Each major piece of equipment is designed to withstand seismic forces of 0.25g horizontally and 0.125g vertically. The ASME Boiler and Pressure Vessel Code, Section VIII-1965, is specified for pump casings, heat exchanger, filter vessels, and the sludge tank, as well as for the fuel pool surge tanks. The fuel pool filters and the surge tanks are shielded with concrete to give a design radiation level of 5 mr/hr outside the shielded area. 3.0 Design Evaluation Precoat-type filters capable of removing particles as small as 1 micron are provided, although experience indicates that 25-micron particle size filtration should be sufficient to maintain pool clarity. Each pump filter heat exchanger loop is adequately sized to handle the normal heat load of the spent fuel storage facility, providing a complete standby loop. The two loops are adequate to handle the full core discharge storage heat load. Various precautions are taken to assure minimum loss of water from the system. All penetrations into the pool are located at a minimum height from the bottom such that there will always be at least 1 ft of water above the fuel. Siphon breakers are used where necessary and the pumps are sealed externally. For flexibility, either pump may be used with a given filter heat exchanger loop. Makeup water to the spent fuel storage pool is provided by the condensate transfer system. The condensate transfer system can be supplied emergency power from the diesel generators, ensuring the supply of makeup water in the event of loss of both normal and reserve ac power. Makeup water is also available to the spent fuel storage pool through the fire protection system by the use of a water hose. NMP Unit 1 UFSAR Section X X-39 Rev. 25, October 2017 The fuel pool cooling system is controlled from a local panel. The Operator is provided with indications of system flow, pool water level, water temperature (on both sides of the heat exchangers), sludge tank level, and valve positions. Alarms are provided on the annunciator and the computer for high- and low-pressure flow and temperature where critical. The spent fuel pool system may be secured for maintenance for limited periods as long as: 1) the time available for the maintenance activity has been predicted by an approved calculation, which ensures the pool temperature will remain below 110°F; 2) the pool temperature is closely monitored during the maintenance activity to ensure the temperature does not exceed 110°F (the maintenance time available may be increased based on this empirical data); and 3) the condensate transfer system is available for makeup. 4.0 Tests and Inspections All equipment in this system will be normally operated, as spent fuel and other components are stored in the pool. However, if equipment such as the spare pump filter heat exchanger loop should stand idle for some time, it will be exercised to assure that it operates properly. I. BREATHING, INSTRUMENT AND SERVICE AIR SYSTEM 1.0 Design Bases A reliable supply of clean, filtered air fit for human breathing is distributed to various areas of the Station. Breathing air is filtered and meets the specifications of ANSI Z86.1 Specification G-7.1 Grade D, 1973. A reliable supply of clean dry air for use of instruments and controls is also provided. Air is supplied at a temperature not exceeding 100°F. The pressure of the instrument air system is controlled between a normal range of 95 and 110 psig, which meets or exceeds the pressure required by some equipment in the Station. A reliable supply of service air is provided for use in maintenance, alarm and trip functions for the preaction and NMP Unit 1 UFSAR Section X X-40 Rev. 25, October 2017 drypipe fire sprinklers, and as a backup for the instrument air system and condensate filtration system (CFS) backwash air. Service air is supplied at a temperature not exceeding 100°F and at a pressure between a normal range of 98 and 105 psig. When the service air compressor is removed from service for maintenance, a portable air compressor can be used to maintain the service air pressure. An air receiver with 2,000 cu ft in volume is provided for testing the containment spray system, as described in Section VII-B. 2.0 System Design The system is shown on Figure X-9. Breathing air is supplied by one single-stage, 200-scfm, motor-belt driven, teflon ring, nonlube piston air compressor. Outside air is drawn through an intake filter from the turbine building roof, compressed, cooled, filtered of dust, and discharged into a 150-ft3 receiver capable of supplying air to 12 men, at the rate of 7.0 cfm per man for 5 min, in the event of compressor failure. The system pressure is maintained between 93 and 105 psig. In the event of failure of the breathing air compressor, breathing air can be supplied from the instrument air system. Breathing air will be supplied by the instrument air receiver passing through a dust filter, a solenoid blocking valve, and a check valve before discharging into the breathing air receiver. The check valve and automatic closure of the solenoid valve, when instrument air receiver pressure is 80 psig or less, prevents discharge of air from the breathing air receiver into the instrument air receiver should the instrument air system pressure fail. In addition, automatic closure of the solenoid valve prevents service air from entering the breathing air system, should the crosstie trip open and the check valve between the crosstie and the instrument air receiver fail to close. Closure of the solenoid valve is alarmed and ample breathing air remains to provide sufficient time to recall persons using breathing air. At the various breathing air outlets, portable or fixed regulating stations are installed to reduce the air pressure from approximately 100 psig by 10-30 psig. Piping for the breathing air is brass and copper to avoid corrosion products. NMP Unit 1 UFSAR Section X X-41 Rev. 25, October 2017 Air for instruments, controls and as a backup to the breathing air system is supplied by one of the three instrument air compressors. There are two 485-scfm flange-mounted, motor-driven, teflon ring, nonlube piston, 2-stage compressors with a 150-cu ft receiver, and one 674-scfm 2-stage, oil-free, rotary screw, water-cooled compressor with a 210-cu ft receiver. The two 485-scfm compressors are separated from the one 674-scfm compressor by a normally open intertie valve, 94-91, located in the 4-in intertie line. Outside air is drawn through separate intake filters for each instrument air compressor, compressed, cooled, and discharged into the receiver. Air from the receiver then passes through drying and filtering equipment to the instruments, controls, and to certain processes requiring high-pressure, oil-free air. This air is available in all buildings and at all levels. Service air for use in maintenance and as a backup to the instrument air system and CFS system air compressor is supplied by one double-stage, nominal 500-scfm, flange-mounted, motor-driven, teflon ring, nonlube piston air compressor. Outside air is drawn through an intake filter from the turbine roof, compressed, cooled and discharged into a 151-cu ft receiver. Air from the receiver then is supplied to outlets maintaining a pressure between 98 and 105 psig not to exceed 100°F. Service air provides backup for the instrument air. A crosstie is located after the instrument air receiver, but before the dryer and filters. It is set to trip open only if the instrument air supply pressure decreases below 90 psig. With the crosstie open, the system will continue to receive air. Check valves located in the crosstie line prevent backfeeding of instrument air into the service air and of service air into the instrument air receiver. The function of the containment spray system air test is covered in Section VII-B, Containment Spray System. During normal operation of the Station, the 2,000-cu ft containment spray system air test receiver is isolated from the containment spray system, and functions as an additional instrument and breathing air receiver. It is capable, together with the 150-cu ft NMP Unit 1 UFSAR Section X X-42 Rev. 25, October 2017 instrument air receiver, of furnishing instrument air for at least 15 min after failure of the instrument air compressors, before air pressure would decrease to 75 psig and service air would be required for backup. 3.0 Design Evaluation Clean dry air is provided in the system design for instrumentation, breathing, and containment spray system testing. The three instrument, one breathing and one service air compressors are of oil-free cylinder construction. Air passing through the instrument air dryers has its dew point lowered to -10°F. Upon exiting the dryer, instrument air passes through either of two parallel filters. The dual parallel filter arrangement allows filter maintenance to be performed during air system operation. Instrument air servicing the waste disposal building and other radwaste systems passes through a refrigerant-type dryer and through either of two parallel filters. Reliable operation of instrument air end users and in-line components is dependent on the filtration and removal of particulates greater than 40 microns. Additional filtration for various components exists where the 40 micron limit is not satisfactory. System reliability is provided by redundancy of compressors, a large receiver system and the service air system crosstie. The two 485-scfm instrument air compressors are each sized to furnish full system requirement on a duty cycle of approximately 75 percent or 85 percent duty cycle including air drying. The 674-scfm instrument air compressor is sized to furnish full system requirements on a duty cycle of approximately 55 percent or 65 percent including air drying. The two 485-scfm compressors are on standby. In the event that the duty compressor fails, one of the standby compressors automatically takes over the load. In addition, the 485-scfm compressors are available for operation in an emergency, since it is possible to operate the compressors and their cooling systems with power from the emergency diesel generators. NMP Unit 1 UFSAR Section X X-43 Rev. 25, October 2017 In the event the piping fails downstream of the 729-scfm instrument air compressor, intertie valve 94-91 will close at approximately 89 psig as sensed in the 4-in intertie line. This will allow the two 485-scfm compressors to continue to supply their loads. The presence of nonsafety-related loads on the safety-related air system does not degrade reliability and performance of the instrument air system. The large receiver capacity and the combination of the control air receiver and the containment spray system air test receiver provides at least 15 min of instrument air at pressures above 75 psig, should all three of the instrument compressors fail. At the conservative setting of 90 psig, the service air system crosstie trips open and the system requirements are provided for by the service air system. The redundancy and reliability provided in this system are necessary since loss of instrument air would necessitate shutdown of the Station. An analysis of the effects of an instrument air failure is given in Section XV, Instrument Air Failure Malfunction Analysis. 4.0 Tests and Inspections Compressor duty will be rotated between compressors on a scheduled basis, providing opportunity to observe the operation and performance of all compressors. Critical temperatures and pressures are continuously monitored and alarmed. Surveillance of the system filters is accomplished by monitoring and alarming the differential pressure across the filters. J. FUEL AND REACTOR COMPONENTS HANDLING SYSTEM 1.0 Design Bases The design of facilities for handling fresh and irradiated fuel and reactor components is based on the following major considerations. 1. Fresh fuel will be received and stored, with or without flow channels, in a manner which precludes inadvertent criticality and at the Nine Mile Point NMP Unit 1 UFSAR Section X X-44 Rev. 25, October 2017 Nuclear Station Independent Spent Fuel Storage Installation (ISFSI). 2. Normal reactor refueling will involve replacement of approximately 35-40 percent of the core. 3. Spent fuel is stored onsite in the spent fuel storage pool until offsite disposal is available. 4. The fuel assemblies and other reactor components to be handled are of the size and weight given in Section IV, Reactor. 5. Previously irradiated flow channels should not be installed on fresh fuel bundles; only new channels should be installed on fresh fuel bundles. Irradiated channels normally are not used on different bundles. 6. There will be no release of contamination or exposure of personnel to radiation in excess of 10CFR20 limits. 7. The spent fuel pool is currently licensed for 4086 spent fuel storage locations using the neutron absorber material Boral, with 1840 storage locations in the north half of the pool and 2246 locations in the south half. Reracking of the north half of the spent fuel pool was completed after the 1999 refuel outage, increasing the north half storage locations to 1840. The reracking of the south half of the pool, increasing from 1710 to 2246 storage locations, has been partially completed. Six of the eight existing Boraflex racks have been replaced with new Boral racks, increasing the capacity from 1296 to 1656 storage locations. Two Boraflex racks remain in the south half, providing 414 storage locations. The rerack of the remaining two racks has been deferred until further capacity increase is warranted. 8. It will be possible at any time to perform limited work on irradiated components. 9. Storage space can be provided for irradiated control rods, flow channels and other reactor components. NMP Unit 1 UFSAR Section X X-45 Rev. 25, October 2017 10. Fuel sipping operations and the temporary installation of fuel sipping equipment. 2.0 System Design 2.1 Description of Facility The major components of this system are a fresh fuel storage vault, spent fuel storage pool, a cask drop protection system, reactor head cavity, reactor internals storage pit, refueling platform, and other auxiliary equipment, all of which are located on the operating floor (see Figure X-10). The fresh fuel vault is a reinforced concrete Class I structure, accessible through top hatches and a personnel door. Racks in the vault can hold a maximum of 200 fuel bundles in an upright attitude. The center-to-center spacing of bundles in the racks is 6.6 in by 11 in. There is an open drain in the floor of the vault. An area monitor used as a criticality monitor is installed in the vault (see Section XII-B). The spent fuel storage pool is a reinforced concrete Class I structure. The interior of the pool is lined with stainless steel plate. Leak detection channels at the liner seams connect to open telltale drains. A stainless steel-lined canal connects the pool to the reactor head cavity for fuel transfer. During normal operation the canal is closed by two sealed aluminum gates in series. An open drain with a flow switch alarm is provided between the gates for leak detection. The depth of water in the pool is 37 ft 10 in. The depth of water in the transfer canal during refueling is 22 ft 9 in. The water in the pool is continuously filtered and cooled by the spent fuel storage pool filtering and cooling system described in Section X-H. There are two types of spent fuel storage racks in the spent fuel storage pool. Both are designed to maintain an adequate criticality margin (Keff less than 0.95) under all storage conditions. Calculational methodology and the treatment of uncertainties and manufacturing tolerances are described in References 1, 2, and 3 for the Boraflex racks and in References 10 and 31 for the Boral racks. The spent fuel storage pool is currently licensed for 4086 fuel assemblies. After the 1999 refuel outage, 1840 storage locations using the neutron absorber NMP Unit 1 UFSAR Section X X-46 Rev. 25, October 2017 material Boral were installed in the north half of the pool. The reracking of the south half of the pool, increasing from 1710 to 2246 storage locations, has been partially completed. Six of the eight existing Boraflex racks have been replaced with new Boral racks, increasing the capacity from 1296 to 1656 storage locations. Two Boraflex racks remain in the south half, providing 414 storage locations. The rerack of the remaining two racks has been deferred until further capacity increase is warranted. Through the use of the neutron absorber material Boral, the high-density racks in the north end of the pool may contain up to 1840 fuel assemblies, and the south end of the pool may contain up to 2246 fuel assemblies. The remaining Boraflex high-density poison racks in the south end of the spent fuel pool may hold up to 414 fuel assemblies with a peak design lattice enrichment of 18.13 grams (3.75 weight percent) of U-235 per axial centimeter. The rows are separated by neutron absorber material--boroflex poison, which is a matrix of boron carbide powder in a silicon polymer. Peak design lattice enrichment is arrived at by averaging the rod enrichments on the design drawings for each lattice, and then choosing the highest (peak) lattice average. Manufacturing tolerances are not included. The Boral racks in the north and south ends of the pool have been analyzed for the design basis fuel assembly of a standard 8x8 array of boiling water reactor (BWR) fuel rods containing UO2 clad in Zircaloy (62 fuel rods with 2 water rods). The GE11 fuel design, a 9x9 array of fuel rods with 8 partial-length fuel rods, and the GNF2 fuel design, a 10x10 array of 92 fuel rods and 2 water holes, have also been analyzed and found to be less reactive for a given enrichment than the GE 8x8 fuel. A maximum k of 1.31 in the standard cold core geometry will encompass the GE11 and GNF2 fuel at all enrichments up to 4.6 weight percent. Therefore, the fuel assemblies stored in these racks must have a peak lattice enrichment of 4.6 percent or less, and the k in the standard cold core geometry, calculated at the maximum over-burnup, must be less than or equal to 1.31. Any fuel of 3.1 percent average enrichment or less is acceptable regardless of the gadolinium content or the k in the standard core geometry. The racks are designed so that accidental dropping of a fuel assembly will not cause critical geometry. Five mountings for jib cranes are installed around the periphery of the pool. Jib cranes with a 1/2-ton capacity can be installed on these mountings for handling components in the pool, and for transferring fresh fuel from the storage vault to the pool. The northwest corner of the pool is reserved for NMP Unit 1 UFSAR Section X X-47 Rev. 25, October 2017 loading spent fuel shipping or transfer casks, and is provided with a cask drop protection system (Section X-J.2.1.1). The cask drop protection system is not required to be used with a dry cask storage transfer cask. The reactor head cavity is completely lined with stainless steel plate. Leak detection channels at each liner seam connect to open telltale drains. A flexible bellows seal connects the liner to the drywell shell. A horizontal steel refueling seal platform inside the drywell shell, and connected to the reactor vessel flange by a second flexible bellows seal, closes off the bottom of the reactor head cavity. During refueling, the cavity is filled with water to a height of 48 ft 9 in above the top of active fuel (TAF). This water is filtered and cooled by the spent fuel storage pool filtering and cooling system. Additional cooling can be provided by the reactor shutdown cooling and cleanup systems. These systems are described in Sections X-H, X-A and X-B, respectively. A second stainless steel-lined transfer canal connects the reactor head cavity to the reactor internals storage pit. During refueling, this canal is filled with water to a depth of 19 ft 6 in. Steel-sheathed concrete plugs, 5 ft 5 in thick, shield the refuel floor from the reactor head cavity during power operation. Both canals are filled with similar plugs during power operations. The concrete plugs provide approximately 4 ft of shielding between the equipment storage pit and the reactor head cavity, and approximately 4.5 ft of shielding between the spent fuel pool and the reactor head cavity. The reactor internals storage pit is a reinforced concrete pit, completely lined with stainless steel. The pit is flooded with water to a depth of 24 ft 0 in during refueling. This water is circulated through the spent fuel storage pool filtering and cooling system. The pit is large enough to accommodate the reactor steam separator and the reactor steam dryer assemblies side by side. The refueling platform is equipped with a 1600-lb capacity main hoist, and two 1,000-lb capacity auxiliary hoists. Each of these hoists can be positioned over any point in the reactor head cavity or the spent fuel storage pool. NMP Unit 1 UFSAR Section X X-48 Rev. 25, October 2017 Protective interlocks, discussed in the Refueling Accident, Section XV, are installed in the power supplies to the refueling platform to prevent inadvertent reactivity additions to the core during refueling. The operating floor is serviced by the reactor building crane, which is equipped with a 125-ton main hoist and a 25-ton auxiliary hoist. These hoists can reach all areas of the operating floor. The 125-ton main hoist is also equipped with a redundant hoisting system, which will prevent the dropping of heavy loads in the event that a cable or other critical part of the main hoist equipment should fail. Three 1/2-ton capacity portable jib cranes are provided for operations in the fresh fuel storage vault and the spent fuel storage pool. Mountings (five in all) for these cranes are provided around the periphery of the pool. A variety of tools for remote handling of fuel and reactor internals and flow channel exchange are provided. Fuel sipping may be required to identify fuel assemblies that contain failed fuel rods. Additionally, the fuel assemblies identified to contain failed rods may be inspected and repaired in the fuel prep machine. In order to perform this work, it is required to store fuel sipping equipment in empty control blade rack cells until such time that the sipping operation is complete. 2.1.1 Cask Drop Protection System The cask drop protection system has been designed to 1) prevent loss of spent fuel pool integrity as a result of certain types of cask drop accidents which may occur over the spent fuel pool, and 2) minimize damage to spent fuel and other components stored in the pool. Specifically, the system has been designed to meet the following functional requirements: 1. Prevent the cask from tipping into the spent fuel pool. 2. Guide the falling cask into the hydraulic dashpot section of the structure. 3. Control the attitude of the cask as it falls through NMP Unit 1 UFSAR Section X X-49 Rev. 25, October 2017 the guide structure and dashpot assembly. 4. Decelerate the cask to a low impact velocity. 5. Absorb the energy of the cask upon impact. 6. Limit loads transmitted to the floor of the spent fuel pool to acceptable values. This system consists of a circular base plate attached to the bottom of the shipping cask and a combination guide structure--dashpot assembly which is permanently installed in the spent fuel pool (Figure X-11). The structural design of the cask drop protection system is based on the worst-case hydraulic, vertical and lateral loadings associated with a wide range of postulated cask drop accidents.(4,5,6) This design provides protection against a wide range of different size and weight shipping casks. A summary description of the basis for conducting safe heavy load movements is provided in Section X-J.2.3. Sufficient protection from the risk associated with potential heavy load drops is also provided by satisfying the guidelines of NUREG-0612, Sections 5.1.1 and 5.3(7,8). 2.2 Operation of the Facility Fresh fuel is brought into the reactor building through the reactor building track bay extension shown on Figure III-4, and hoisted to the operating floor through the equipment hatch utilizing the reactor building crane. (See Figures III-5 to III-9.) The fresh fuel is removed from its shipping containers, inspected, flow channels attached, and stored in the fresh fuel storage vault. Normally prior to refueling, the fresh fuel is transferred to the spent fuel storage pool using the 25-ton auxiliary overhead hoist. In preparation for refueling, the concrete shield plugs in the reactor head cavity and the transfer canals are removed by the reactor building crane. The drywell head and reactor vessel head are removed using the same crane. NMP Unit 1 UFSAR Section X X-50 Rev. 25, October 2017 The steam dryer and the steam separator assemblies are transferred to the reactor internals storage pit. Water levels are controlled such that the steam separator is transferred submerged. During the disassembly process, demineralized condensate is pumped into the reactor until the head cavity and the reactor internals storage pit are flooded to the normal level of the spent fuel storage pool. The spent fuel storage pool gates are removed after the water level has reached the normal level of the spent fuel storage pool. Spent fuel is removed from the reactor using a grapple attached to the refueling platform and placed in racks in the spent fuel storage pool. The same equipment is used to transfer the fuel from the spent fuel storage pool to the reactor. At the completion of reactor refueling, the moisture separator, steam dryer and reactor head are put back into place following the proper maintenance procedures. The drywell head and concrete shield blocks are then restored. After refueling, the spent fuel bundles are stored in spent fuel storage pool racks. They will remain there until NRC/DOE resolution of disposal problems is finalized or the spent fuel pool is placed into dry cask storage at the onsite NMPNS ISFSI. 2.3 Control of Heavy Loads Program 2.3.1 Introduction/Licensing Background NUREG-0612 provides regulatory guidelines for the control of heavy loads to assure the safe handling of heavy loads in areas where a load drop could impact stored spent fuel, fuel in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. In a letter dated December 22, 1980(11) (later identified as GL 80-113), as supplemented by GL 81-07(12) and GL 83-42(20), the NRC requested that licensees describe how these guidelines were satisfied at their facility. This request was divided into two phases (Phase I and Phase II). The Niagara Mohawk Power Corporation (NMPC) response to the Phase I portion of the request for Unit 1, addressing the guidelines of Section 5.1.1 of NUREG-0612, was initially provided in letters dated May 22, July 28, and NMP Unit 1 UFSAR Section X X-51 Rev. 25, October 2017 September 22, 1981(13-15). Supplemental information was subsequently provided in NMPC letters dated August 1, 1982; September 30, November 15, and December 15, 1983; July 26, 1984; and January 18, August 5, and November 25, 1985(16-19,21,22,24,25). By letter dated March 5, 1985(7), the NRC issued their safety evaluation which concluded that the guidelines in NUREG-0612, Sections 5.1.1 and 5.3, had been satisfied for Unit 1, and that Phase I of the NMPC response for Unit 1 was acceptable. In GL 85-11(8), the NRC documented their determination that a detailed Phase II review of heavy loads was not necessary and that Phase II was considered completed. By letter dated May 13, 1996(27), NMPC provided the required response to Bulletin 96-02(26) for Unit 1. The response reiterated that the movement of heavy loads over critical areas of the refuel floor and safety-related equipment is performed in accordance with controlled site procedures developed in accordance with NUREG-0612. Additionally, the response reaffirmed that the reactor building 125-ton crane is single-failure-proof (i.e., a dual load path, redundant hoisting system). The NRC's April 23, 1998, letter(28) accepted the NMPC response and indicated completion of tasks associated with Bulletin 96-02. On July 28, 2008(29), the Nuclear Energy Institute (NEI) transmitted NEI 08-05, Revision 0, Industry Initiative on Control of Heavy Loads to the NRC. This document was issued to provide an industry agreed-upon approach to providing additional assurance of compliance to existing regulatory guidelines regarding control of heavy loads at nuclear power plants. This NEI document includes guidance associated with updating UFSARs to reflect a summary description of the basis for conducting safe heavy load movements. The NRC safety evaluation of the guidelines contained in NEI 08-05, transmitted to NEI by letter dated September 5, 2008(30), determined that the guidelines may be used by licensees to establish a revised licensing basis for handling of reactor vessel heads and other heavy loads, subject to the clarifications and conditions noted in the NRC's safety evaluation. 2.3.2 Safety Basis NMP Unit 1 UFSAR Section X X-52 Rev. 25, October 2017 Heavy load handling activities pose a safety risk in the areas of nuclear power plants where load drops could impact irradiated fuel or equipment necessary for safe shutdown. Implementing the guidelines of NUREG-0612, Section 5.1.1, reduces the potential for heavy load drops and provides a measure of defense-in-depth against such an occurrence. The risk associated with load handling failures is acceptably low based on meeting the Phase I requirements of NUREG-0612, Section 5.1.1, and the use of the reactor building 125-ton single-failure-proof crane for lifting the reactor vessel head and spent fuel casks. The 125-ton reactor building crane is a single-failure-proof crane as defined in NUREG-0612, Appendix C, and has a redundant hoisting system which is independently capable of supporting the crane's rated load(23). 2.3.3 Scope of Heavy Load Handling Systems In NUREG-0612, the scope of cranes includes: "Overhead handling systems that are used to handle heavy loads in the area of the reactor vessel or spent fuel in the spent fuel pool. Additionally, loads may be handled in other areas where their accidental drop may damage safe shutdown systems..." Based on the NMPC Phase I responses in References 13 through 19, 21, 22, 24, and 25, the reactor building 125-ton crane is within the scope of Section 5.1.1 of NUREG-0612. Heavy load movements in areas of safe shutdown equipment that are handled by other load-handling systems are also performed in accordance with the requirements of NUREG-0612 as defined in controlled site procedures. These other systems include, but are not limited to, the reactor building 25-ton auxiliary crane, the turbine building 150-ton crane, and the screen and pump house 25-ton crane. 2.3.4 Control of Heavy Loads Program The Control of Heavy Loads Program consists of the following: 1. NMPNS commitments in response to NUREG-0612, Phase I elements, as described in References 13 through 19, 21, 22, 24, and 25. NMP Unit 1 UFSAR Section X X-53 Rev. 25, October 2017 2. For reactor pressure vessel head (RPVH) and spent fuel cask lifts, the single-failure-proof reactor building 125-ton crane described in Section X-J.2.3.4.2 is used. 2.3.4.1 NMPNS Commitments in Response to NUREG-0612, Phase I Elements NMPNS has committed to controlling the movement of heavy loads in accordance with the seven elements of Section 5.1.1 of NUREG-0612, as defined below: 1. Safe load paths for movement of heavy loads are defined in controlled plant procedures(14). 2. Controlled plant procedures are developed and implemented that control movement of heavy loads(14). 3. Crane operators are trained and qualified in accordance with controlled plant procedures(14). 4. Special lifting devices follow the guidelines of ANSI N14.6-1978(15). 5. Lifting devices not specifically designed follow the guidelines of ANSI B30.9-1971(15). ASME B30.9-2010 is used for the selection, use and maintenance of synthetic round slings. 6. The reactor building 125-ton crane is inspected, tested and maintained consistent with ANSI B30.2-1976(14). 7. The reactor building 125-ton crane is designed to CMAA-70 and meets the applicable criteria and guidelines of ANSI B30.2-1976(14,15). 2.3.4.2 Reactor Pressure Vessel Head and Spent Fuel Cask Lifts The reactor building 125-ton crane is single-failure-proof, has a redundant load path, and is designed to CMAA-70. The following attributes were defined in the design of the crane(23): NMP Unit 1 UFSAR Section X X-54 Rev. 25, October 2017 1. Allowable stress limits are defined and conservative enough to prevent permanent deformation of individual structural members when exposed to maximum load lifts. 2. The crane is capable of stopping and holding the load during a design basis earthquake. 3. Automatic controls and limiting devices are designed so that they fail-safe and do not prevent the crane from stopping and holding the load safely. 4. The design of the wire rope reeving system includes dual wire ropes. 5. Limit switches are included to limit such items as overspeed, overload and overtravel and cause the hoisting action to stop when limits are exceeded. 6. The reeving system is designed against the destructive effects of "two-blocking." 7. Safety devices such as limit switches are provided to reduce the likelihood of a malfunction. 2.3.5 Safety Evaluation Controls implemented by NUREG-0612, Phase I elements, together with the use of a single-failure-proof crane for RPVH and spent fuel cask lifts, make the risk of a load drop extremely unlikely and acceptably low. The risk associated with the movement of heavy loads is evaluated and controlled by station procedures. 3.0 Design Evaluation The spacing of fuel bundles in the fresh fuel storage vault maintains keff <0.95 even if flooded with water. The vault floor drain prevents flooding. The spacing of fuel bundles in the spent fuel storage pool maintains keff <0.95. A criticality monitor in the fresh fuel storage vault provides warning in the unlikely event of a criticality incident. Protective interlocks prevent handling of fuel over the reactor when a control rod is withdrawn. Another set of interlocks prevents control rod withdrawal when fuel is being handled over NMP Unit 1 UFSAR Section X X-55 Rev. 25, October 2017 the reactor. Limit switches on the refueling platform hoists interrupt power to the hoists when the TAF is 8 ft below the surface of the water. Brakes on all equipment lock upon loss of power. Spent fuel will not be inadvertently handled with an inadequate depth of water shielding. The above interlocks can be bypassed to permit the unloading of a significant portion of the reactor core (full core offload, spiral offload) for such purposes as removal of temporary control curtains, CRD maintenance, inservice inspection (ISI) requirements, examination of the core support plate, etc. (Technical Specification 3.5.3). Fuel stored in the spent fuel storage pool is covered by a minimum of 24 ft of water. Irradiated fuel being moved is at all times covered by a minimum depth of 8 ft of water over TAF, except that the fuel preparation machine is provided with mechanical stops to ensure that active fuel remains under 7 ft of water. Spent fuel pool water level is automatically controlled to ensure that during normal operation, spent fuel will be covered by a sufficient depth of water to permit unrestricted access to the operating floor. The spent fuel storage pool cannot be completely drained. If draining should be initiated due to Operator error, level alarms will notify operating personnel and makeup water will be supplied automatically. If no action were taken, the fuel would still be covered by approximately 1 ft of water after the pool had drained down to the lowest penetration. All reactor servicing operations are carried out within the secondary containment, which is described in Section VI-C. A bypass around the refueling platform radiation monitor will allow the monitor to be connected into the RPS during refueling operations or when recently irradiated fuel or a fuel-loaded shipping or transfer cask is being handled. This monitor provides a fast automatic isolation of the reactor building ventilation system and initiation of the reactor building emergency ventilation system. 4.0 Tests and Inspections During testing prior to initial reactor fueling, the spent fuel storage pool, reactor head cavity, and reactor internals storage pit were filled with water and checked for leakage. Dummy fuel NMP Unit 1 UFSAR Section X X-56 Rev. 25, October 2017 assemblies were run through a complete cycle from the fresh fuel storage vault to the spent fuel storage pool. During normal operation, telltales are examined for evidence of potential leakage from the spent fuel pool. Prior to fuel handling, all hoists, cranes and tools are inspected and tested to assure safe operation. K. FIRE PROTECTION PROGRAM The fire protection program is based on the NRC requirements and guidelines, Nuclear Electric Insurance Limited (NEIL) Property Loss Prevention Standards and related industry standards. With regard to NRC criteria, the fire protection program meets the requirements of 10 CFR 50.48(c), which endorses, with exceptions, the Nation-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants 2001 Edition. Nine Mile Point Nuclear Station Unit 1 has further used the guidance of NEI 04-for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR -Informed, Performance Fire Protection for Existing Light-Water Nuclear Adoption of NFPA rformance-Based Standard for Fire 2001 Edition in accordance with 10 CFR 50.48(c) serves as the method of satisfying 10 CFR 50.48(a) and General Design Criterion 3. Prior to adoption of NFPA 805, General Design Criterion 3, "Fire Protection" of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utilization Facilities," was followed in the design of safety and non-safety related structures, systems, and components, as required by 10 CFR 50.48(a). NFPA 805 does not supersede the requirements of GDC 3, 10 CFR 50.48(a), or 10 CFR 50.48(f). Those regulatory requirements continue to apply. However, under NFPA 805, the means by which GDC 3 or 10 CFR 50.48(a) requirements are met may be different than under 10 CFR 50.48(b). Specifically, whereas GDC 3 refers to SSCs important to safety, NFPA 805 identifies fire protection systems and features required to meet the NMP Unit 1 UFSAR Section X X-57 Rev. 25, October 2017 Chapter 1 performance criteria through the methodology in Chapter 4 of NFPA 805. Also, under NFPA 805, the 10 CFR 50.48(a)(2)(iii) requirement to limit fire damage to SSCs important to safety so that the capability to safely shut down the plant is satisfied by meeting the performance criteria in Section 1.5.1 of NFPA 805. A Safety Evaluation was issued on June 30, 2014 by the NRC, that transitioned the existing fire protection program to a risk-informed, performance-based program based on NFPA 805, in accordance with 10 CFR 50.48(c). Design Basis Summary 1.1 Defense-in-Depth The fire protection program is focused on protecting the safety of the public, the environment, and plant personnel from a plant fire and its potential effect on safe reactor operations. The fire protection program is based on the concept of defense-in-depth. Defense-in-depth shall be achieved when an adequate balance of each of the following elements is provided: (1) Preventing fires from starting, (2) Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage, (3) Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed. 1.2 NFPA 805 Performance Criteria The design basis for the fire protection program is based on the following nuclear safety and radiological release performance criteria contained in Section 1.5 of NFPA 805: Nuclear Safety Performance Criteria. Fire protection features shall be capable of providing reasonable assurance that, in NMP Unit 1 UFSAR Section X X-58 Rev. 25, October 2017 the event of a fire, the plant is not placed in an unrecoverable condition. To demonstrate this, the following performance criteria shall be met. (a) Reactivity Control. Reactivity control shall be capable of inserting negative reactivity to achieve and maintain subcritical conditions. Negative reactivity inserting shall occur rapidly enough such that fuel design limits are not exceeded. (b) Inventory and Pressure Control. With fuel in the reactor vessel, head on and tensioned, inventory and pressure control shall be capable of maintaining or rapidly restoring reactor water level above top of active fuel for a BWR such that fuel clad damage as a result of a fire is prevented. (c) Decay Heat Removal. Decay heat removal shall be capable of removing sufficient heat from the reactor core or spent fuel such that fuel is maintained in a safe and stable condition. (d) Vital Auxiliaries. Vital auxiliaries shall be capable of providing the necessary auxiliary support equipment and systems to assure that the systems required under (a), (b), (c), and (e) are capable of performing their required nuclear safety function. (e) Process Monitoring. Process monitoring shall be capable of providing the necessary indication to assure the criteria addressed in (a) through (d) have been achieved and are being maintained. Radioactive Release Performance Criteria. Radiation release to any unrestricted area due to the direct effects of fire suppression activities (but not involving fuel damage) shall be as low as reasonably achievable and shall not exceed applicable 10 CFR, Part 20, Limits. Chapter 2 of NFPA 805 establishes the process for demonstrating compliance with NFPA 805. NMP Unit 1 UFSAR Section X X-59 Rev. 25, October 2017 Chapter 3 of NFPA 805 contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection systems and features. Chapter 4 of NFPA 805 establishes the methodology to determine the fire protection systems and features required to achieve the nuclear safety performance criteria outlined above. The methodology shall be permitted to be either deterministic or performance-ce criteria, defense-in-depth, and safety margin and require no further engineering analysis. Once a determination has been made that a fire protection system or feature is required to achieve the nuclear safety performance criteria of Section 1.5, its design and qualification shall meet the applicable requirement of Chapter 3. 1.3 Codes of Record The codes, standards and guidelines used for the design and installation of plant fire protection systems are as follows: (for specific applications and evaluations of codes refer to appropriate upper tier document(s)) NMP Unit 1 UFSAR Section X X-60 Rev, 25 October 2017 DEVIATIONS FROM NFPA STANDARDS NFPA STANDARD SECTION DEVIATION/JUSTIFICATION NFPA 10 3-1.2.1 Deviation: Complete protection of buildings with Class A extinguishers not provided. Justification: Class A units are provided where Class A materials are present; balance of Station covered by Class B-C units. NFPA 13 (also applies to NFPA 14, 15, 16 & 24) 2-7.1 Deviation: Fire department connections are not provided at Unit 1. Justification: Backup for site pumps is provided by interconnection with Unit 2. City supply provides limited capability. The Unit 1 site (150 psi) pump pressures and volume are sufficient without the use of fire department pumpers. NFPA 13 (also applies to NFPA 15 & 16) 3-17.6 Deviation: The normal solenoid valves used on the listed Viking deluge valves have been replaced with MOVs. Justification: There was concern for accidental system operation of premature shutdown in the event of loss of power to the valves. Dual MOVs powered by separated circuitry were therefore employed. NFPA 13, 1980 Edition 4-2.1.4 Deviation: Several sprinklers on preaction sprinkler system WP-4116, reactor building, el 281 ft, north side, are approximately 11 ft from the primary containment wall. NFPA 13, 1980 edition, paragraph 4-2.1.4, states that the maximum distance from a wall should not exceed 1/2 the maximum distance allowed between sprinklers. The sprinkler system was designed to protect an Ordinary Hazard occupancy. According to paragraph 4-2.1.2, the maximum allowable spacing between sprinklers for an Ordinary Hazard occupancy is 15 ft; therefore, the maximum allowable distance from a wall is 7 1/2 ft. Justification: Preaction sprinkler system WP-4116 is designed to control fires occurring in various reactor building fire break zones (FBZ) and to prevent any fire occurring on one side of a zone from crossing to the other side. The sprinklers are spaced such that they provide full area coverage to the floor where any combustibles might be located. The sprinkler spacing does not exceed the 7 1/2 ft requirement if taken from the base of the primary containment wall. The code is written for typical construction and anticipates vertical walls, while the walls of the primary containment slope away from the floor. Therefore, while the spacing meets the requirement at the base of the wall, it does not meet the requirement at the level of NMP Unit 1 UFSAR Section X X-61 Rev, 25 October 2017 Table (Cont'd.) NFPA STANDARD SECTION DEVIATION/JUSTIFICATION the sprinkler system piping. The existing sprinkler spacing would, however, provide sufficient water spray into the area above the sloping wall to prevent a fire from extending from one side of the FBZ to the other, as there are extremely limited combustibles in this space. Although the system will provide adequate protection for the hazard, the sprinkler spacing is technically in deviation to NFPA 13, 1980 edition, paragraph 4-2.1.4. NFPA 13, 1996 Edition (see FPEE 1-98-004, Rev. 1 for reasoning on code of record year) 4-5.1 Deviation: Sprinklers shall be positioned and located so as to provide protection of an area consistent with the overall objectives of NFPA 13. The sprinklers (the existing foam-water nozzles which now only discharge water) beneath the condensers are approximately 18 in above the condenser pit floor. NFPA 13, Figure 4-5.5.1.1 shows that at 18 in a sprinkler spray pattern only covers a floor area of 50 sq ft as the water will only spray a maximum of 4 ft radially at this height. Sprinklers beneath the condenser are spaced approximately 14 ft on the branch lines. This creates 6-ft gaps between spray patterns. Water should be discharged (distributed) over the entire surface of an oil pool fire to provide extinguishment. Justification: Even with degraded protection, a fire in the condenser pit would be contained (although not fully suppressed) within the pit by the existing suppression system. Also, fire fighting plans were upgraded to include the manual application of foam. See FPEE 1-98-004, Rev. 1 for further details. NFPA 13, 1996 Edition (see FPEE 1-98-004, Rev. 1 for reasoning on code of record year) 4-5.5.3 Deviation: Sprinklers shall be installed under horizontal obstructions that interfere with sprinkler spray patterns. There are no sprinklers beneath the open Justification: The reason for this requirement is to prevent an excessive amount of sprinklers from opening during a fire. Opening an excessive amount of sprinklers is not an issue in this area as the suppression system is deluge type where all the sprinklers are already open. In addition, the deluge system provides a large area cooling effect that acts to provide exposure protection to structures and equipment even if spray patterns are partially interrupted by the mezzanine. See FPEE 1-98-004, Rev. 1 for further details. NMP Unit 1 UFSAR Section X X-62 Rev, 25 October 2017 Table (Cont'd.) NFPA STANDARD SECTION DEVIATION/JUSTIFICATION NFPA 14 4-2.2 Deviation: Pressure limiting devices are not installed to limit standpipe outlet pressure to 100 psig. Justification: Fire hoses connected to the standpipe system are intended for use exclusively by the site Fire Brigade. All Fire Brigade members receive hands-on training in the proper handling of high-pressure fire hose. This training reduces the potential for high-pressure hose to be a hazard to the user. NFPA 14 7-7.1 Deviation: Installed pressure gauges are not provided at the top of each standpipe. Justification: Pressure readings on all standpipe risers can be taken with a portable gauge. NFPA 15 (also applies to NFPA 13) 4-4.1.4 Deviation: The cable tray sprinkler system employs design elements of NFPA 13 and 15. Justification: NFPA 15 specifies coverage requirements for cable trays using open nozzles. NFPA 13 deals with area protection using fused sprinklers or nozzles. Elements of both standards were employed for design as they applied. Fused nozzles were chosen to limit potential water damage to equipment. NFPA 20 7-1.1.1 7-6 Deviation: The electric fire pump is powered by a 4-kV motor which is controlled by an unlisted circuit breaker. Justification: NFPA 20 permits use of high-voltage motor starters, none of which are UL listed. The similar starter was evaluated by a UL engineer at Unit 2 to ensure compliance with the requirements of NFPA 20. NMP Unit 1 UFSAR Section X X-63 Rev, 25 October 2017 Table (Cont'd.) NFPA STANDARD SECTION DEVIATION/JUSTIFICATION NFPA 20 8-2.1.1 Deviation: The diesel fire pump is not listed for fire service by an approved testing laboratory. Justification: The authority having jurisdiction at the time (NMPC) accepted the manufacture and installation of the diesel fire pump as having met the intent of National Board of Fire Underwriters (NBFU) Standard #20, Centrifugal Fire Pumps, and Underwriters Laboratories Approval Listing Requirement UL-448, Pumping Equipment for Private Fire Service. NFPA 20-1980 8-6.1 Deviation: NFPA 20 requires that the electric motor-driven fire pump be tested weekly. Testing of the electric motor-driven fire pump for operational readiness is performed every 31 days. Justification: NFPA inspection, testing, and maintenance requirements are intended to cover a broad group of users. Nuclear plants operate under unique conditions that inherently foster high reliability. These favorable operating conditions are conducive to performance-based analysis methods that provide quantitative evidence of high system reliability. NFPA has developed a performance-based approach to fire protection at nuclear power plants. This approach recognizes these unique features and allows for changes and deviations from the normal code requirements. Based on analysis using performance-based techniques, NEIL has recognized that nuclear power plants can be considered outside the normal NFPA code guidelines and has developed their own interpretation of the testing requirements to satisfy their insurance requirements. NEIL recommends testing electric motor-driven fire pumps on a less frequent basis than that recommended by NFPA (monthly vs. weekly). Performing a monthly test frequency for verifying the operational readiness of the electric motor-driven fire pump maintains the licensing and property insurance requirements, and will provide adequate verification of operational readiness of this fire pump. NMP Unit 1 UFSAR Section X X-64 Rev, 25 October 2017 Table (Cont'd.) NFPA STANDARD SECTION DEVIATION/JUSTIFICATION NFPA 20 9-5.2.7 Deviation: The diesel-driven fire pump is not equipped with a weekly program timer. Justification: Station procedure requires a 30-min operating test of the pump weekly. NFPA 72-2002 10.4.3 Deviation: Fire detectors are not tested annually. Fire detectors are demonstrated operable in accordance with the following test methodology: at least 10 percent of the installed detectors, with a minimum of one detector in each detection loop, shall be tested annually by initiating an alarm per the methods described in the surveillance procedures. Should a detector fail to alarm under test conditions, it will be corrected per procedure, and an additional 20 percent, with a minimum of two detectors in the affected loop, shall be tested. Should a failure to alarm occur in this expanded sample population, the failure will be corrected per procedure, and all remaining detectors in the affected loop will be tested and corrected as necessary. This testing methodology will be cycled through all detectors in a detection loop until all detectors in the loop have been tested. All detectors in a loop shall be tested within a 10-yr time frame. The cycle will then be repeated. Where detector testing cannot be accomplished within the time period specified by the surveillance procedures, due to accessibility or safety concerns during plant operation, the testing for those detectors shall be performed during the next cold shutdown exceeding 24 hr. NMP Unit 1 UFSAR Section X X-65 Rev, 25 October 2017 Table (Cont'd.) NFPA STANDARD SECTION DEVIATION/JUSTIFICATION Justification: Based on plant-specific fire detector failure-to-alarm data documented in FPEE-1-97-002 and DER 2001-4956, the above methodology provides an equivalent level of safety as that required by NFPA 72-2002, and is in accordance with the process for approving alternate methods of protection, as outlined in NFPA 72-2002, paragraphs 1.2.3, 1.5.1, and 1.5.2. NFPA 72-2002 edition was used as the basis for code compliance, as the most up-to-date code is typically utilized when reassessing test requirements. NFPA 72D 2-2.2 Deviation: The control room and LFCPs are not UL listed. Justification: The Unit 1 panels are designed to meet the operational needs of the Station, which results in the need for substantially larger-than-normal enclosures. UL-listed enclosures which satisfy these requirements are not available; therefore, nonlisted enclosures are used. The panel components are UL listed. NFPA 72D 3-3.1 Deviation: Manual-pull fire alarm boxes are not used in the Station. Justification: The Unit 1 Gaitronics two-way public address system is designed and installed to provide fire reporting capability in accordance with the intent of NFPA 72D. NFPA 72E 4-3.7 Deviation: For structural conditions involving beams in excess of 18 in deep and spaced less than 8 ft apart, smoke detectors are mounted on the bottom flange of alternate beams. Justification: NFPA 72E is silent on the acceptable coverage under the structural configuration noted. The chosen arrangement was reviewed with the 72E Chapter 4 Subcommittee Chairman and found acceptable. NMP Unit 1 UFSAR Section X X-66 Rev, 25 October 2017 Table (Cont'd.) NFPA STANDARD SECTION DEVIATION/JUSTIFICATION NFPA 80 2-5.4 Deviation: Hollow metal swinging fire doors are required to have a door/frame clearance of not more than 1/8 in for headjam, sides, and between double doors. Maximum clearance at door bottom is 3/8 in when there is a noncombustible raised sill and 3/4 in where no sill is present. These clearances may be increased by an additional 1/8 in provided a fire protection engineering evaluation (FPEE) is prepared that gives adequate justification (which includes that the required minimum latch engagement into the strike is satisfied). Note: Required minimum latch throw is either stamped on door label or obtained from NFPA 80 Table 2-8B. Required minimum latch throw minus 1/8 in (the maximum NFPA door clearance) equals the required minimum latch engagement. The actual measured latch throw minus the actual door clearance must match or exceed required minimum latch engagement. Justification: Fire tests performed for other utilities indicate that fire doors with door gaps 1/8-in larger than the NFPA 80 limit still pass the standard 3-hr fire test. Many fire doors would not be exposed to a 2- or 3-hr duration fire due to: low fire loading on either side of the door; the presence of fixed suppression systems; the presence of smoke/heat detection signaling systems which would initiate an immediate fire brigade response to a fire in its early stages. The major fire door requirement is that it stay in place without opening during the fire. Fire doors can warp when exposed to a fire. This warpage can cause the latch bolt to retract from the strike plate, allowing the door to swing open. Therefore, the required minimum latch engagement into the strike must be maintained. The possibility of additional smoke penetration past the door (due to slightly larger door clearances) hampering egress is not considered a concern at Unit 1 since: the vast majority of the plant is not normally occupied; normally-occupied plant areas have a low occupant load; most fire areas have at least two means of egress. NMP Unit 1 UFSAR Section X X-67 Rev, 25 October 2017 Table (Cont'd.) NFPA STANDARD SECTION DEVIATION/JUSTIFICATION NFPA 90A 2-1.4.1 Deviation: Service openings are not provided adjacent to duct-installed smoke detectors. Justification: Duct-type smoke detectors used can be serviced from the outside of the duct. NFPA 90A 2-1.4.3 Deviation: Service openings are not provided at 20-ft intervals along the duct and at the base of vertical risers. Justification: Service openings are provided where required for system maintenance. NFPA 90A 3-3.1 Deviation: Labeled fire dampers do not exist at penetrations provided for the reactor building ventilation supply and return air ducts. Justification: Penetrations at these locations are ASTM-36 steel pipe, 54-in diameter. Closure of penetrations is provided by double isolation valves of heavy steel and cast iron construction. Current configuration evaluated to ample duct protection. NFPA 90A 4-3 Deviation: HVAC systems fans are not arranged to shut down automatically on fire detector operations. Justification: The necessity to initiate HVAC system isolation would be evaluated by operations personnel following consideration of location, number of alarms and impacts to Station operations. NMP Unit 1 UFSAR Section X X-68 Rev. 25, October 2017 2.0 System Description 2.1 Required Systems Nuclear Safety Capability Systems, Equipment, and Cables Section 2.4.2 of NFPA 805 defines the methodology for performing the nuclear safety capability assessment. The systems equipment and cables required for the nuclear safety capability assessment are contained in DCD-805. Fire Protection Systems and Features Chapter 3 of NFPA 805 contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection systems and features. Compliance with Chapter 3 is documented in DCD-805. Chapter 4 of NFPA 805 establishes the methodology and criteria to determine the fire protection systems and features required to achieve the nuclear safety performance criteria of Section 1.5 of NFPA 805. These fire protection systems and features shall meet the applicable requirements of NFPA 805 Chapter 3. These fire protection systems and features are documented in DCD-805. Radioactive Release Structures, systems, and components relied upon to meet the radioactive release criteria are documented in DCD-805. 2.2 nuclear plant operations. For the purposes of establishing the structures included in the fire protection program in accordance with 10 CFR 50.48(c) and NFPA 805, the plant structures listed Power Block Structures Fire Area(s) Reactor Building FA1, FA2, FA3 Administration Building FA4, FA12 Turbine Building FA5, FA6, FA7, FA9, FA16A, FA16B, FA17A, FA17B NMP Unit 1 UFSAR Section X X-69 Rev. 25, October 2017 Power Block Structures Fire Area(s) Control Complex FA10, FA11 Diesel Generator Building FA18, FA19, FA20, FA21, FA22, FA23, FA24 Screenhouse FA13, FA14 Radwaste Storage and Solidification Building FA15 Waste Building FA15 Offgas Building FA5 Yard (115kV Switchyard, 345kV Switchyard, West End of Unit 1 (includes HH2, SWG#1 & 2 and duct bank, H2 tank), and Transformer Area (includes transformers XF-101S, XF-101N, XF-TB01, XF-TB02, and XF-T10)) EXT 3.0 Safety Evaluation The DCD-805 documents the achievement of the nuclear safety and radioactive release performance criteria of NFPA 805 as required by 10 CFR 50.48(c). This document fulfills the requirements of of NFPA 805. The document contains the following: Identification of significant fire hazards in the fire area. This is based on NFPA 805 approach to analyze the plant from an ignition source and fuel package perspective. Summary of the Nuclear Safety Capability Assessment (at power and non-power) compliance strategies. o Deterministic compliance strategies o Performance-based compliance strategies (including defense-in-depth and safety margin) Summary of the Non-Power Operations Modes compliance strategies. Summary of the Radioactive Release compliance strategies. NMP Unit 1 UFSAR Section X X-70 Rev. 25, October 2017 Summary of the Fire Probabilistic Risk Assessments. Key analysis assumptions to be included in the NFPA 805 monitoring program. 4.0 Fire Protection Program Documentation, Configuration Control and Quality Assurance In accordance with Chapter 3 of NFPA 805 a fire protection plan documented in DCD-805 defines the management policy and program direction and defines the responsibilities of those individuals -805: Designates the senior management position with immediate authority and responsibility for the fire protection program. Designates a position responsible for the daily administration and coordination of the fire protection program and its implementation. Defines the fire protection interfaces with other organizations and assigns responsibilities for the coordination of activities. Identifies the appropriate authority having jurisdiction for the various areas of the fire protection program. Identifies the procedures established for the implementation of the fire protection program, including the post-transition change process and the fire protection monitoring program. Identifies the qualifications required for various fire protection program personnel. Identifies the quality requirements of Chapter 2 of NFPA 805. Detailed compliance with the programmatic requirements of Chapters 2 and 3 of NFPA 805 are contained in DCD-805. L. REMOTE SHUTDOWN SYSTEM 1.0 Design Bases NMP Unit 1 UFSAR Section X X-71 Rev. 25, October 2017 This system is designed to provide the plant Operators with hot shutdown capability independent of the main and auxiliary control rooms. The remote shutdown system (RSS) was designed to safely affect hot shutdown during a fire which causes a functional loss and/or evacuation of the main and auxiliary control rooms. 2.0 System Design The RSS consists of two independent panels. Each panel controls one set of emergency condenser return isolation, inlet isolation and makeup level control valves. Reactor scram is achieved by de-energizing MG set 131 from one panel and MG set 141 from the other panel. Parameters monitored at each remote shutdown panel (RSP) include analog indications for: Reactor pressure Reactor water temperature Reactor water level Torus water temperature Drywell pressure Emergency condenser water level PB 102 and 103 voltage Drywell temperature A set of status lights is provided to indicate: All control rods in Reserve power transfer control circuits position Emergency condenser return isolation valve position Emergency condenser inlet isolation valve position Emergency condenser makeup level control valve position 3.0 System Evaluation Since each panel only de-energizes one MG set, both panels must be manned to achieve full scram. Only one channel of either RSP monitoring instrument or control will be required to bring the reactor safely to hot shutdown after a full scram has been achieved. Each panel is located in a separate fire area, relative to each other and the main and auxiliary control room. The shutdown panels will be activated when the control room is uninhabitable. Manual transfer switches which are individually fused provide power to the RSP. The isolation switch at each NMP Unit 1 UFSAR Section X X-72 Rev. 25, October 2017 RSP is keylocked. An alarm is annunciated in the main control room when the isolation switch is placed in the remote position (this position can override control room operation). In addition, access to each shutdown panel is controlled by a keylocked cage. 4.0 Tests and Inspections Tests are performed to assure the operability of the RSPs and the reliability of the transfer switches to effectively override the control room from each panel. M. HYDROGEN WATER CHEMISTRY AND NOBLE METAL CHEMICAL ADDITION (NOBLECHEM) SYSTEMS 1.0 Design Basis The HWC and NMCA systems are provided to mitigate intergranular stress corrosion cracking (IGSCC) of the recirculation piping and the reactor vessel internals. Mitigation of IGSCC in operating boiling water reactors (BWR) can be effectively accomplished by reducing the bulk liquid oxidant (oxygen and hydrogen peroxide). Hydrogen added to the feedwater suppresses the radiolytic generated oxidant concentration in the core regions, and enhances the recombination reactions in the downcomer. The reduction in oxidant level can reduce the ECP significantly and crack initiation and growth also are greatly reduced, even at high bulk liquid oxidant levels. Reducing the ECP requires high hydrogen addition rates which result in increased main steam line radiation levels from volatile 16N compounds. The catalytic behavior of noble metals provides an opportunity to efficiently achieve a dramatic reduction in ECP by catalytically reacting hydrogen with all oxidants at the catalytic surface. NobleChem employs the reactor coolant as the transport medium to deposit minute amounts of noble metal on all wetted reactor components. With the ratio of hydrogen to oxygen in excess of stoichiometric, the corrosion potential of the reactor vessel and internal components decreases significantly, and crack initiation and growth also are greatly reduced, even at high bulk liquid oxidant levels. Low hydrogen addition rates are still necessary to provide sufficient excess hydrogen at the surface of NobleChem treated components. Oxidants that diffuse to the component surface will immediately react with the excess hydrogen (molar ratio of hydrogen to oxidant >2) to form water. In this way, the NMP Unit 1 UFSAR Section X X-73 Rev. 25, October 2017 boundary layer of all NobleChem wetted components is depleted of oxidants and a very low corrosion potential is maintained. In summary, NobleChem utilizes very reactive surfaces to maintain oxidant deficient water in contact with reactor components. Therefore, because of the lower operational dose rates, the NobleChem process in conjunction with low hydrogen addition rates is an effective approach to mitigate and prevent IGSCC. 1.1 Noble Metal Chemical Addition System The NMCA process involves periodic injection of noble metal compounds using either the classic method or the on-line method. Classic Hot Shutdown Application The classic method involves periodic injection of noble metal compounds, containing platinum (Pt) and rhodium (RH), into the recirculation loop(s) and into the reactor vessel, through existing small bore piping connections in the recirculation pump differential pressure transmitter lines. The noble metal compounds are deposited on reactor internal surfaces with the reactor in hot standby condition. The noble metal compounds are distributed by circulating coolant using 3 of 5 recirculation pumps. The resulting coolant flow across the core and core shroud is relatively uniform enhancing proper deposition on wetted surfaces. Appropriate water level in the reactor vessel is maintained by operating the CRD and the RWCU systems. Normal reactor coolant makeup is available per operations procedures for this hot shutdown condition. The classic noble metal deposition process lasts approximately 48 hr, with the coolant temperature maintained between 250°F and 350°F as required by the General Electric-Nuclear Energy (GENE) Application Procedure. The exact temperature during the application within this range is a GENE process decision, as is the rate of chemical injection. During the process period, a combination of the recirculation pumps and shutdown cooling is used to regulate the coolant temperature. On-Line Power Operations Application (OLNC) The OLNC process only injects the platinum compound [NA2Pt(OH)6] into the reactor vessel through the feedwater system during power operation. To get sufficient catalyst into the cracks and crevices, and stay within nominal chemistry control limits during normal reactor operation, the rate of Pt injection must be significantly reduced and the period of injection increased NMP Unit 1 UFSAR Section X X-74 Rev. 25, October 2017 as compared to the classic NobleChem process (which has injection rates as high as 40g per hour). As a minimum, a typical time period for on-line application is expected to be about 2 weeks (and injection rates less than one-tenth of the classic process). The noble metal compounds are deposited onto all surfaces that come into contact with the moving reactor coolant in the applicable temperature range. For example, at a nominal deposition of 1µg/cm2, the uniform coverage is approximately one atom layer of 3 Å thickness (1 Å is 1 x 10-7 mm or 3.94 x 10-9 in). Surface scans of autoclave treated specimens have shown that the noble metal atoms present on the surface do not completely cover the surface, but are distributed randomly across the surface. On an atomic scale, the deposited noble metals are discontinuous. Even with agglomeration, the maximum thickness of Pt and Rh is significantly less than 0.001 in, which is less than the minimum manufacturing tolerances of the vessel components (e.g., the tolerance of the fuel Zircaloy tubes is 0.003 in and the Zircaloy channels is 0.004 in). 1.2 Hydrogen Water Chemistry System The HWC system injects hydrogen into the feedwater system at the suction to the feedwater booster pumps. The injected hydrogen causes a reduction in dissolved oxygen within the reactor internals and recirculation piping and lowers the radiolytic production of hydrogen and oxygen in the vessel core region. Hydrogen addition to the feedwater results in an excess ratio of hydrogen to oxygen at the entrance to the offgas (OFG) system. Therefore, the HWC system also provides an oxygen supply upstream of the OFG recombiner to maintain stoichiometric mixture of hydrogen and oxygen in the recombiner. With the suppression of radiolysis, the main steam line dissolved oxygen concentration decreases. This can result in lower condensate and feedwater dissolved oxygen concentrations. If the condensate and feedwater dissolved oxygen concentration values are less than 20 ppb, accelerated carbon steel corrosion can occur. Unit 1 has an existing oxygen injection system to add oxygen to the condensate feedwater system. The HWC system has a provision to supply oxygen to supplement the existing oxygen injection system for the condensate feedwater system. 2.0 System Design 2.1 Noble Metal Chemical Addition NMP Unit 1 UFSAR Section X X-75 Rev. 25, October 2017 The NMCA system consists of temporary skid-mounted injection equipment and permanently installed monitoring equipment. The monitoring equipment includes two ECP monitoring locations, material coupons at the durability monitor for monitoring the noble chemistry coating thickness, the capability for crack growth rate monitoring, and a sample supply for the NMCA temporary equipment needed for the injection of the NMCA material. The major components of the NobleChem application system include:

• NobleChem injection system
• Sampling systems
• NobleChem analysis system
• Data acquisition system The NobleChem injection system is composed of an injection subsystem and a drive water subsystem which, respectively, control the injection rate of the noble metal solutions and provide transport of the chemicals into the vessel. The injection and drive water subsystems are mounted on individual portable skids. One portable skid-mounted sampling system provides the ability to obtain liquid samples for measuring conductivity and pH of the reactor water during the application process. In addition, grab samples can be collected for analysis to determine the noble metal concentrations in the reactor water. Also, the sample skid will include a material deposition sampler that will provide tubing samples for monitoring the application process. The NobleChem analysis system provides the ability to determine the noble metal concentrations in the reactor water during the application process. The process utilizes an inductively-coupled plasma spectrometer with a secondary mass spectrometer. Specified ion concentrations also can be determined through the use of an ion chromatograph. A special NobleChem DAS records the data collected during the application process. 2.2 Hydrogen Water Chemistry System The HWC system consists of the flow monitoring and control equipment for both hydrogen and oxygen, a hydrogen isolation module, a system control module, and an offgas oxygen monitor NMP Unit 1 UFSAR Section X X-76 Rev. 25, October 2017 panel. The control room interface includes a shutdown switch, status lights and annunciators. 2.2.1 HWC Feedwater Hydrogen Injection The new feedwater hydrogen injection system is designed to deliver, meter and inject hydrogen into the feedwater system at the suction side of the feedwater booster pumps. The system produces a reduction in dissolved oxygen within the reactor internal components and recirculation piping, and reduces the radiolytic production of hydrogen and oxygen in the vessel core region. The system includes a hydrogen injection control panel and the hydrogen isolation module. New piping is installed to deliver hydrogen gas to the panels and to the injection points. The panel/module regulates the quantity of hydrogen injection into the feedwater system. 2.2.2 HWC Offgas Oxygen Injection Hydrogen addition to the feedwater system results in an excess ratio of hydrogen to oxygen at the entrance to the OFG system. Therefore, the HWC system also provides an oxygen supply upstream of the offgas recombiner to maintain a stoichiometric mixture of hydrogen and oxygen in the recombiner. The HWC OFG oxygen injection system includes piping to deliver oxygen and an oxygen injection control panel. The main HWC panel/module (PNL500) regulates the injection of oxygen into the OFG system. The system also is designed to provide oxygen to the existing Unit 1 condensate oxygen injection system that is used for maintaining proper condensate water chemistry. This feature of the new HWC oxygen injection system replaces the existing stand-alone oxygen gas bottle source. 2.2.3 HWC Offgas Sample The HWC addition includes an additional offgas monitoring panel for monitoring of the offgas percent oxygen concentration from the recombiners to assure that the oxygen addition and hydrogen addition flows are properly balanced. The HWC OFG sample system draws gas from downstream of the offgas vent coolers. 3.0 System Evaluation The HWC and NMCA systems are classified as nonsafety-related systems. The basic function of the systems is to create an environment in the reactor coolant system (RCS) that will NMP Unit 1 UFSAR Section X X-77 Rev. 25, October 2017 minimize IGSCC of stainless steel components in the reactor coolant recirculation piping and lower reactor internals. The systems are not required for safe shutdown of the plant and are not required to mitigate the consequences of an accident. The systems are designed to be safe and reliable, consistent with the requirements of using hydrogen and oxygen gases inside the plant. 4.0 Tests and Inspections The preoperational test is performed to confirm the operability of the installed system components and piping configurations. The startup test is performed to verify the function of all system components and the capability of the system to control the process. N. REFERENCES 1. Application for Amendment to Operating License, Docket No. 50-220, from D. P. Dise (NMPC), dated March 21, 1978. 2. Letter from C. V. Mangan (NMPC) to Domenic B. Vassallo (NRC), dated June 24, 1983, Spent Fuel Pool Modification. 3. Letter from C. V. Mangan (NMPC) to Domenic B. Vassallo (NRC), dated October 5, 1983, Spent Fuel Storage Capacity Expansion. 4. Letter from D. P. Dise (NMPC) to V. Stello (NRC), July 17, 1978. (Attached information: Movement of Heavy Loads Near the Spent Fuel Storage Pool or Reactor Core). 5. Letter from T. J. Brosnan (NMPC) to J. F. O'Leary (AEC), September 29, 1972. (Attached report: NMP-1 Cask Drop Protection System). 6. Letter from P. D. Raymond to D. L. Zieman (AEC), May 31, 1973. (Attached report: NMP-1 Cask Drop Protection System). 7. Letter from D. B. Vassallo (NRC) to B. G. Hooten (NMPC), March 5, 1985, Control of Heavy Loads (Phase 1). 8. NRC Generic Letter 85-11, June 28, 1985, Completion of Phase II of "Control of Heavy Loads at Nuclear Power Plants," NUREG-0612.

NMP Unit 1 UFSAR Section X X-78 Rev. 25, October 2017 9. DRFC 11-61, "Calculation of Free Volume Available for Scram with Leakage for SDV Long-Term," dated June 1, 1981. 10. Technical Specification Amendment No. 167 for NMP1 to Reflect a Planned Modification to Increase the Storage Capacity of the Spent Fuel Pool, dated June 17, 1999. 11. Letter from D. G. Eisenhut (NRC) to All Licensees, December 22, 1980, Control of Heavy Loads (Generic Letter 80-113). 12. Letter from D. G. Eisenhut (NRC) to All Licensees, February 3, 1983, Control of Heavy Loads (Generic Letter 81-07). 13. Letter from D. P. Dise (NMPC) to D. G. Eisenhut (NRC), May 22, 1981. 14. Letter from D. P. Dise (NMPC) to D. G. Eisenhut (NRC), July 28, 1981. 15. Letter from D. P. Dise (NMPC) to D. G. Eisenhut (NRC), September 22, 1981. 16. Letter from T. E. Lempges (NMPC) to D. G. Eisenhut (NRC), August 1, 1982. 17. Letter from T. E. Lempges (NMPC) to D. G. Eisenhut (NRC), September 30, 1983. 18. Letter from T. E. Lempges (NMPC) to D. G. Eisenhut (NRC), November 15, 1983. 19. Letter from T. E. Lempges (NMPC) to D. G. Eisenhut (NRC), December 15, 1983. 20. Letter from D. G. Eisenhut (NRC) to All Licensees, December 19, 1983, Clarification to Generic Letter 81-07 Regarding Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" (Generic Letter 83-42). 21. Letter from C. V. Mangan (NMPC) to D. B. Vassallo (NRC), July 26, 1984. 22. Letter from T. E. Lempges (NMPC) to D. B. Vassallo (NRC), January 18, 1985. NMP Unit 1 UFSAR Section X X-79 Rev. 25, October 2017 23. Letter from P. D. Raymond (NMPC) to D. J. Skovholt (NRC), July 26, 1973. 24. Letter from T. E. Lempges (NMPC) to D. B. Vassallo (NRC), August 5, 1985. 25. Letter from C. V. Mangan (NMPC) to J. A. Zwolinski (NRC), November 25, 1985. 26. NRC Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core or Over Safety Related Equipment, April 11, 1996. 27. Letter from C. D. Terry (NMPC) to U. S. Nuclear Regulatory Commission, May 13, 1996, NRC Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core or Over Safety Related Equipment. 28. Letter from D. S. Hood (NRC) to J. H. Mueller (NMPC), April 23, 1998, Completion of Licensing Action for NRC Bulletin 96-02 for Nine Mile Point Nuclear Station, Unit Numbers 1 and 2. 29. Letter from A. R. Pietrangelo (NEI) to E. J. Leeds (NRC), July 28, 2008, NEI 08-05, Revision 0, Industry Initiative on Control of Heavy Loads. 30. Letter from W. H. Ruland (NRC) to T. C. Houghton (NEI), September 5, 2008, Industry Initiative on Control of Heavy Loads. 31. S22.1-XX-G079NF (Holtec HI-2114838), "Fuel Criticality Analysis for GNF2 Fuel in Boral," dated March 21, 2011. 32. Letter from Bhalchandra Vaidya (NRC) to C. Costanzo dated June 30, 2014 - Nine Mile Point Nuclear Station, Unit 1 Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c). REACTOR SHUTDOWN COOLllG SYSTEM FIEACTOO BUILOIHG CLOSED LOOP coo..:t£ *.o:irEFi srs1s-1 REF. FIG.. X-4 [;J I ---liEUEF TOREACtat 8Uti..0111G EoutPIEl'IT DAAlll TN*: §-,, ,..-.. P:JMP5TMT rs nnEl'l.CO: [;'.] E!TI---*>-+----+J R:LEF TO ltE$CTCA llUILDlllG EOUll't.iEltT OAA!lolTNlo' %--G OR1'WELL REACTOR SYSTE:M >------: \ -----8 --B , .. "'-, I WAT'i:.'R SEAL 1-=FtOM CORE SPRAY iTSTE'.11 re --G FIGURE X-1 UFSAR Re vision 22 O:tober 2011 REACTOR CLEAN-ILJP SYSTEH I I I I i I PU>P ElECTRltAL. CDfiTRtl.. CACUlT TO WAS Tl O!Sl"OSAL SY5TEtt FIGURE x-:-* UFSAR R. O . e v1s1on :::.::: c-r.ober* 2011 L '""""'" Of.Ml!EIW.IZEll CONTROL ROD DRIVE HYDRAULIC SYSTEM IEHIEJIG(loC,C(IN()(N5£R CCIUNSA1* 1CtEP41.U REACTOR .......... r-'""""--"""---+----'----------.--.....!.--ElttW.1.iT WllTERIEllmlt COOL .... WllEAl*Ml£A P"-"""'----'--'-----+--'----'----DPIYE WAftJI HIE"'(l(lt FIGURE X-3 UFSAR Rev1s10n 21 October 200'l I ¢ I I I I I $ I I I I I I I I $ I I I I I I I I I I I I I I ::0 rn D n ___, 0 ::0 CTJ c I 0 z Cl n I D (fl rn 0 I 0 D u n D D I z Cl (fl -< (fl ___, rn :s: " REf.FlG.:<-6 TURBINE BLJILDfNG CLOSED LOOP COOLING SYSTEM .. NOTE I

• SHAfT DAl\'Et-1 REACTOR FEEO\oiAlER Pl.Mf COOLERS TL.RB!llE MUXIL If<R!E!i1 CL CC-OLERS STE>IM PACt-:H.IG EXHAUSTER CCO...ER.5 MECHA"HCAL C.OOLERS CiEf.E;RATOR HYDROGEN COOLERS COQERS GaiERA '!"OR LEAO COO LEAS iLRa:NE BUiLDING EO'J£Pt-Ef.lf OOHHI Hlfa\ *11 trJSTRUMElff AIR COMPRESSOR 0"5TRUHENT [jfrfEA *12 .. 2
• f£CfRCiJ>.HTWC Plrlf Nt!fGA GEr*ERATOR SET COCl.Ef\S HOUSE SERYICE MR SA!fl.E COOLERS BATTERY ROOM 14 HJR l:tl'IDI flOl'lER Tl.ftlUIJI!: 111.JIL.DINO C:LDSEC LL'.UCJl'.IUl'l:W4TEJ!l'V4PS 8 I I ' ----+---8 l I I *--ffi--1..--B EJ:ERCiCICl' Jld.'.E\Jf'TD ""'-'""" FIGURE X-5 liCll!lTNG Cl.C'itD:...D:I' ctlOL..ltC UFSAR Re,11s1on 22 October 2011

..... ..... SERVICE WATER SYSTEM PAGE X-32 FIGURE X-6 UFSAR Rev1s1on 25 OCTOBER 2flll7 x6.dgn Nine Mile Point Unit 1 UFSAR FIGURE X-7 THIS FIGURE HAS BEEN DELETED UFSAR Revision 17 October 2001 SPENT FUEL STORAGE POOL FIL TERJNG AND CODLING SYSTEM r .. fl L V ---!>l---l><J--r==r--1 '"'---0 l I i 0 g] 0 g] g] . MAKE-LP OEMINERALIZER CLOSE.0 LOOP COOLING EQUIPMENT CLEANUP SYSlEM EQUIPMENT " FUEL POOL SYSTEM EQUIPMENT ' A -FUEL POO\. SYSTEM EQUIPMENT -REACTOR BUILDING SEAL WATER TURBINE BUILDING SEAL WATER RESIN TRANSFER ANO CLEANING EOUJPMENT C1Q RAOWASTE EOOIPMENT -OFF-OAS EQUIPMENT FEEDWATER EQUIPMENT -CJ) . A --M ! u (-j \7 -'ii "' t n 0 z 0 I'll z (/) J> -I I'll -I :0 J> z (/) "Tl I'll :0 (/) -< (/) -I I'll 3: FEEDWATER FLOW SYSTEM txl I IXI s !Xi--1><1--""""""" lill!:ClllCIUTl.CHft.IN\ FIGURE XI -7 UFSAR Revision 24 October 2015 U.S. NUCLEAR REGULATORY COMMISSION DOCKET 50-220 LICENSE DPR-63 NINE MILE POINT NUCLEAR STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT (UPDATED) OCTOBER 2017 REVISION 25 NMP Unit 1 UFSAR LIST OF EFFECTIVE FIGURES SECTION XII Figure Revision Number Number Section XII EF XII-1 Rev. 25, October 2017 XII-1 20 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section XII XII-i Rev. 25, October 2017 SECTION XII RADIOLOGICAL CONTROLS A. RADIOACTIVE WASTES 1.0 Design Bases 1.1 Objectives 1.2 Types of Radioactive Wastes 1.2.1 Gaseous Waste 1.2.2 Liquid Wastes 1.2.3 Solid Wastes 2.0 System Design and Evaluation 2.1 Gaseous Waste System 2.1.1 Offgas System 2.1.2 Steam-Packing Exhausting System 2.1.3 Building Ventilation Systems 2.1.4 Stack 2.2 Liquid Waste System 2.2.1 Liquid Waste Handling Processes 2.2.2 Sampling and Monitoring Liquid Wastes 2.2.3 Liquid Waste Equipment Arrangement 2.2.4 Liquid Radioactive Waste System Control 2.3 Solid Waste System 2.3.1 Solid Waste Handling Processes 2.3.2 Solid Waste System Equipment 2.3.3 Process Control Program 3.0 Safety Limits 4.0 Tests and Inspections 4.1 Waste Process Systems 4.2 Filters 4.3 Effluent Monitors 4.3.1 Offgas and Stack Monitors 4.3.2 Liquid Waste Effluent Monitor B. RADIATION PROTECTION 1.0 Primary and Secondary Shielding 1.1 Design Bases 1.2 Design 1.2.1 Reactor Shield Wall 1.2.2 Biological Shield 1.2.3 Miscellaneous 1.3 Evaluation 2.0 Area Radioactivity Monitoring Systems NMP Unit 1 UFSAR Section Title Section XII XII-ii Rev. 25, October 2017 2.1 Area Radiation Monitoring System 2.1.1 Design Bases 2.1.2 Design 2.1.3 Evaluation 2.2 Area Air Contamination Monitoring System 2.2.1 Design Bases 2.2.2 Design 2.2.3 Evaluation 3.0 Radiation Protection 3.1 Facilities 3.1.1 Laboratory, Counting Room and Calibration Facilities 3.1.2 Change Room and Shower Facilities 3.1.3 Personnel Decontamination Facility 3.1.4 Tool and Equipment Decontamination Facility 3.2 Radiation Control 3.2.1 Shielding 3.2.2 Access Control 3.3 Contamination Control 3.3.1 Facility Contamination Control 3.3.2 Personnel Contamination Control 3.3.3 Airborne Contamination Control 3.4 Personnel Dose Determinations 3.4.1 Radiation Dose 3.5 Radiation Protection Instrumentation 3.5.1 Counting Room Instrumentation 3.5.2 Portable Radiation Instrumentation 3.5.3 Air Sampling Instrumentation 3.5.4 Personnel Monitoring Instruments 3.5.5 Emergency Instrumentation 4.0 Tests and Inspections 4.1 Shielding 4.2 Area Radiation Monitors 4.3 Area Air Contamination Monitors 4.4 Radiation Protection Facilities 4.4.1 Ventilation Air Flows 4.4.2 Instrument Calibration Well Shielding 4.5 Radiation Protection Instrumentation NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section XII XII-iii Rev. 25, October 2017 XII-1 FLOWS AND ACTIVITIES OF MAJOR SOURCES OF GASEOUS ACTIVITY XII-2 QUANTITIES AND ACTIVITIES OF LIQUID RADIOACTIVE WASTES XII-3 ANNUAL SOLID WASTE ACCUMULATION AND ACTIVITY XII-4 LIQUID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS XII-5 SOLID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS XII-6 OCCUPANCY TIMES XII-7 GAMMA ENERGY GROUPS XII-8 AREA RADIATION MONITOR DETECTOR LOCATIONS NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section XII XII-iv Rev. 25, October 2017 XII-1 RADIOACTIVE WASTE DISPOSAL SYSTEM NMP Unit 1 UFSAR Section XII XII-1 Rev. 25, October 2017 SECTION XII RADIOLOGICAL CONTROLS A. RADIOACTIVE WASTES 1.0 Design Bases 1.1 Objectives The radioactive waste handling systems have been designed to meet the following objectives: 1. Collect and process all radioactive waste generated in the Station without limiting normal Station operation. 2. Collect and process radioactive wastes for disposal, or transfer to a vendor for processing and disposal. 3. Release radioactive material to the environment in a controlled manner so that all releases are within the standards set forth in 10CFR20 and the Technical Specifications. 4. Retain radioactive wastes, if they accidentally leak from the systems, so that they can be recovered and reprocessed. 1.2 Types of Radioactive Wastes 1.2.1 Gaseous Waste Gaseous radioactive wastes include airborne particulates as well as gases vented from process equipment. Sources of gaseous waste activity are the offgas (OFG) system effluent, steam-packing exhauster system effluent, and building ventilation exhausts. Flows and associated activities for the major sources of gaseous activity are given in Table XII-1 for the normal operating condition. Station gaseous discharge limits and atmospheric dispersion rates (see Technical Specifications and Offsite Dose Calculation Manual) limit exposures in the uncontrolled environment to values within the standards given in 10CFR20. NMP Unit 1 UFSAR Section XII XII-2 Rev. 25, October 2017 1.2.2 Liquid Wastes Liquid radioactive wastes include all liquids collected in equipment drains and floor drains in areas of the Station which are potentially contaminated with radioactive materials. Inaddition, shower wastes, laboratory wastes, and decontamination area wastes are handled by the liquid waste system. Flows and associated activities for the major sources of liquid wastes are given in Table XII-2. Liquid wastes are handled in one of the four handling processes described in Section XII-A.2.2. Waste which is discharged to the environment in the cooling water effluent is dispersed in that effluent so that activities in the uncontrolled environment are within the standards listed in 10CFR20 and the Technical Specifications. 1.2.3 Solid Wastes Solid wastes include filter sludge, spent resin, spent condensate filter media, condensate filter backwash sludge, radioactive tools and equipment, and miscellaneous trash from plant operations, laboratory, maintenance and cleanup operations. The solid waste handling system is capable of collecting, processing and temporarily storing these various wastes. Annual accumulation and average activities of these wastes are given in Table XII-3. Solid waste is stored in the waste handling facility for decay and for accumulation of enough waste for shipment to a processor or authorized burial site. Radiation levels of shipped containers are maintained within the standards set forth by the Nuclear Regulatory Commission (NRC) and the Department of Transportation (DOT). 2.0 System Design and Evaluation 2.1 Gaseous Waste System The gaseous waste system is composed of eight major parts. 1. Offgas system NMP Unit 1 UFSAR Section XII XII-3 Rev. 25, October 2017 2. Steam-packing exhauster system 3. Turbine building ventilation system 4. Reactor building ventilation system 5. Waste building ventilation system 6. Stack 7. Offgas building ventilation system 8. Radwaste solidification and storage building (RSSB) ventilation system 2.1.1 Offgas System For a description of the OFG system, see Section XI-B.3. 2.1.2 Steam-Packing Exhausting System A greater volume of gases is handled by this system than by the OFG system. This larger volume of gases results from the addition of room air to the steam leaking from the turbine gland seals. This system is described in detail in Section XI-B.1. 2.1.3 Building Ventilation Systems These systems are described in other sections of the report. 1. Turbine building - Section III-A.2.2 2. Reactor building - Section VI-E.2.0 3. Waste building - Section III-C.2.2 4. RSSB - Sections III-I.1.4 and III-I.2.2 Particulate airborne activity exhausted by each of these systems can be monitored by a constant air monitor (see Section XII-B.2.2). In areas where significant quantities of airborne particulates can be generated, such as the radiochemical laboratory hoods, filters are installed in the exhaust duct to remove these particulates. Because the many tanks and equipment in the waste NMP Unit 1 UFSAR Section XII XII-4 Rev. 25, October 2017 building can be a source of airborne particulate activity, this entire building exhaust is filtered before discharge to the stack. 2.1.4 Stack The stack is described in Section III-G. The stack monitoring system (see Section VIII-C.3.0) is provided to continuously measure the gaseous activity discharged from the stack. This system also incorporates a composite collection of particulate and halogen activity. These filter samples will be removed periodically and the particulate and halogen activity determined in the Station laboratory. The design features of the stack assure that diffusion of the emitted plume will not be significantly influenced by the eddy currents around the Station structures. 2.2 Liquid Waste System 2.2.1 Liquid Waste Handling Processes The liquid waste system is designed to handle four types of liquid waste: high-conductivity waste, low-conductivity waste, chemical waste, and miscellaneous waste. Figure XII-1 is a schematic flow diagram for the liquid waste system and shows the processes for handling all four types of liquid waste. The process for handling each type of waste is described below. 1. High-Conductivity Waste High-conductivity liquid wastes are collected in the floor drain sumps located within the drywell, the reactor building, the turbine building, the RSSB, the offgas building, and the waste disposal building. The wastes in these floor drain sumps are pumped into the floor drain collector, waste neutralizer tank (WNT), or utility collector tank (during power operation and/or when drywell is inerted, the drywell discharge is routed to the waste collector system), which are located in the waste disposal building. After sufficient waste is collected in the floor drain collector, the waste is pumped to one of two floor drain sample tanks and is available for processing. High-conductivity waste from the condensate pre-filter backwash receiver tank (BWRT) is pumped directly to NMP Unit 1 UFSAR Section XII XII-5 Rev. 25, October 2017 the WNT during off-normal operation; e.g., when concentrated waste tank #13 is unavailable. Waste collected in the WNT or utility collector tank may be processed directly from that tank or pumped to the floor drain sample tanks. Waste from either the floor drain collector or utility collector may be processed via the floor drain filter or a combination of the floor drain filter and waste demineralizer for processing through the low conductivity system. High conductivity/low purity aqueous radwaste from the floor drain system is also processed through a series of water treatment modules which include charcoal filtration, small particulate filtration, and demineralization, and, depending on water quality, may include reverse osmosis or ultrafiltration, deionization, oxidizing agents and/or ultraviolet radiation prior to demineralization. Liquid wastes processed in this manner meet the chemistry criteria for recycling to the plant. This water may be directed to the low conductivity/high purity reclamation system (waste collector system) or to the waste sample tanks for chemistry sampling prior to batch transfer to the condensate storage tanks (CST). If the modular processed effluent water quality is unsatisfactory, a conductivity cell will direct the water back to the floor drain system. An alternate processing route for high conductivity liquid waste is the waste concentrator. The distillate from the concentrator is normally recycled to the plant through the waste collector system. Under certain conditions, the liquid waste can be pumped into the circulating water discharge tunnel at a flow rate which will assure that concentrations in the effluent will not exceed Station limits. 2. Low-Conductivity Waste Low-conductivity liquid wastes which usually come from piping and equipment drains are collected in equipment drain sumps or equipment drain tanks located in the drywell, the reactor building, the turbine building, and the waste disposal building. These liquids are pumped to the waste collector tank which is located in the waste disposal building. Other (less frequent) NMP Unit 1 UFSAR Section XII XII-6 Rev. 25, October 2017 sources of low-conductivity waste are waste effluents from the fuel pool cooling system, the reactor cleanup system, the containment suppression chamber, ECs, resin transfer system, the backwash water from the condensate demineralizers (CND), and the clean backwash water from the condensate filtration system (CFS). This waste is also pumped to the waste collector tank in the waste disposal building. A waste surge tank, located in the turbine building, is provided to collect the water from Station system surges and provide interim storage for liquids which may be off-standard and which must be recycled through the liquid waste processing system. The liquid wastes in either the waste collector tank or the waste surge tank are pumped through a high-efficiency precoat type filter and a mixed-bed waste demineralizer to either one of two waste sample tanks. The floor drain filter is also used as a spare filter. Low-conductivity/high-purity aqueous radwaste from the equipment drain/waste collector system may also be processed through a series of water treatment modules which may include charcoal filtration, small particulate filtration, and demineralization, and, depending on water quality, may include reverse osmosis or ultrafiltration, deionization, oxidizing agents and/or ultraviolet radiation prior to demineralization. Liquid wastes processed in this manner meet the chemistry criteria for recycling to the plant and may be directed to the waste sample tanks. While one of the two waste sample tanks is being filled, the other can be sampled, and after sample analysis, the liquid is normally pumped to the CST in the turbine building. Under certain conditions, this liquid can be pumped into the discharge tunnel, after careful analysis, to assure that concentrations in the effluent will not exceed Station limits. In addition to being able to pump fuel pool water and reactor cleanup system water to the waste collector tank, these liquids can also be discharged through filters into the condenser hotwell. From the hotwell, the water is processed through the CNDs and pumped to the CST. NMP Unit 1 UFSAR Section XII XII-7 Rev. 25, October 2017 3. Chemical Waste Chemical waste originates at the chemical addition tank, in the laboratory sinks, and equipment decontamination drains. Since this waste is not only high-conductivity waste but also may contain acids and other chemicals, it is collected in the waste neutralizer tank or utility collector tank in the waste disposal building. The wastes are then neutralized and processed with other high-conductivity waste. 4. Miscellaneous Liquid Waste Liquid waste from the shower facility, personnel decontamination, or any other radioactive liquid waste which might contain detergents, is collected in the waste neutralizer tank, floor drain collector tank, or utility collector tank in the waste disposal building. The waste is then processed with other high-conductivity waste. 2.2.2 Sampling and Monitoring Liquid Wastes Sampling lines are provided from each collection tank and each sample tank, which may be used to evaluate filter and demineralizer performance. These sample lines run to a sample station adjacent to the waste disposal facility control room. In addition, local sample points have been provided, where deemed necessary, throughout the waste facility. Samples are analyzed in the Station laboratory. A composite sample of the circulating water discharge stream is taken at a point downstream of the waste effluent discharge. An aliquot of this composite sample is periodically analyzed in the Station laboratory. Data from samples taken of the tank to be discharged are recorded along with discharge water volume data so that a continuous record is maintained of released activity. The liquid waste discharged is also monitored by a shielded scintillation detector mounted on a vertical section of the liquid waste discharge pipe which leads to the circulating water discharge tunnel. (See Section VIII-C.3.0.) The activity NMP Unit 1 UFSAR Section XII XII-8 Rev. 25, October 2017 detected is recorded and annunciated at abnormally high concentrations, but the activity is still below discharge limits. 2.2.3 Liquid Waste Equipment Arrangement Equipment is arranged and shielded to permit operation, inspections and maintenance with minimum personnel radiation exposure. (Shielding is designed to meet the requirements of Table XII-6.) Sumps, tanks, pumps, instruments and valves are arranged in accessible areas or shielded rooms which are accessible if the equipment is isolated. Process equipment is designed for long life, ease of decontamination, and minimum maintenance. Major equipment and their respective capacities are listed in Table XII-4. The liquid waste disposal system is adequately designed to maintain absolute control over all liquid wastes within the Station during all modes of normal operation. Sufficient reserve tank holdup capacity is provided for the expected short-term high usage conditions. 2.2.4 Liquid Radioactive Waste System Control The liquid waste processing systems are manually initiated and the process controlled from the radioactive waste control panels located in the waste disposal building. Instrumentation and alarms are provided for the control of the process and for detection and signaling of abnormal conditions. In some cases, hose connections and chemical transfer hoses are used to transfer water, water/resin slurry, filter sludge slurry, or water/flocculent slurry. The interconnection of systems or vendor-supplied processing equipment is evaluated and approved prior to use for changes to plant operation and procedures. Administrative procedures are currently used to provide control for use, testing and inspection of chemical transfer hoses. A possible failure of chemical transfer hoses is bounded by the tank rupture failure. Activity discharged to the environment is kept to a practical minimum by the treatment and recycle of much of the waste within the Station, by filtration of much of the waste before NMP Unit 1 UFSAR Section XII XII-9 Rev. 25, October 2017 discharge, and by the concentration of radiological waste and conversion of the resulting concentrate into solid waste. Protection against accidental discharge is accomplished by procedural control, by providing double waste discharge line valves, blanked off lines and locked valves, waste discharge line valve leakoff indication and suitable monitors and alarms of abnormal conditions. To prevent spread of radioactivity within the Station or to outside areas, the significant concentrators of radioactive materials have double-barrier protection. This feature is provided by enclosure within the buildings, containment in concrete pipe tunnels, containment in double-wall piping, and containment in steel sump liners. Consequently, in the event of leaks, spills, and overflows from equipment, containment of the liquid radioactive waste is assured. If a tank is accidentally ruptured, its contents will be contained within the restricted area of the waste disposal building. In some cases, a retention curb is built around a local area or equipment cubicle to contain minor amounts of leakage within the area. The dispersed liquid can be readily recovered and processed as required through the floor drain system. Sumps and drains are provided for the collection and return of wastes to the system. 2.3 Solid Waste System 2.3.1 Solid Waste Handling Processes The solid waste system is designed to handle spent resins, filter sludge, and concentrated waste, as well as to provide for collection and shipment of low-level solids. Wastes may be processed or solidified onsite, or transferred to a vendor for processing. 1. Concentrated Liquid Waste Liquid waste is concentrated and then transferred to an approved container. The containers are then either placed in storage for future shipment, shipped offsite for disposal, or shipped to a vendor for further processing and disposal. 2. Spent Resins and Filter Sludge NMP Unit 1 UFSAR Section XII XII-10 Rev. 25, October 2017 Filter sludges are normally stored in tanks in the buildings where the filters are located. After accumulation, these sludges are transferred as a liquid slurry to the waste building for processing. Spent resins from all demineralizers in the Station, when discarded, are stored in the spent resin storage tank in the waste disposal building. The filter sludge can be processed by routing it through the phase separator system or processing it directly to an approved container. The material is then dewatered or solidified with an approved media, and the container is either placed in storage for future shipment, shipped offsite for disposal, or shipped to a vendor for further processing and disposal. Spent resins are pumped to an approved container in the truck bay where the container is dewatered. The container is then sealed and either placed in storage for future shipment, shipped offsite for disposal, or shipped to a vendor for further processing and disposal. 3. Solid Waste Low-level solid wastes are collected and placed in approved containers. These solids may be compacted or shipped to vendor facilities for volume reduction and disposal or recycling. Special containers may be used for large or odd-shaped components. 4. Miscellaneous Solid Wastes Solid materials such as spent fuel assemblies, spent control blades, poison curtains, in-core chambers, and other equipment originating from the reactor primary system are stored in the spent fuel storage pool until offsite storage or disposal is necessary. Spent fuel assemblies are also stored at the onsite NMPNS ISFSI in dry cask storage. 5. Condensate Filtration System Backwash Waste NMP Unit 1 UFSAR Section XII XII-11 Rev. 25, October 2017 Liquid waste produced by the CFS backwash is treated by the addition of polyelectrolytes. The polyelectrolytes cause the fine iron to settle to the bottom of the converted concentrated waste tank (CWT) #13. The remaining waste water is then decanted and processed by the liquid waste handling system. The remaining sludge is held in the tank until it is transferred by the sludge pump to a disposal liner for disposal. 2.3.2 Solid Waste System Equipment Equipment is arranged and shielded to permit operation, inspections and maintenance with minimum personnel radiation exposure. (Shielding is designed to meet the requirements of Table XII-6.) Highly radioactive wastes are loaded into containers with remotely-controlled equipment and using remote viewing devices. Control of the radwaste system is from the radwaste building control panel or the RSSB control room. Instrumentation is provided both for process control and for detection of abnormal conditions. Major equipment and their respective capacities are listed in Table XII-5. 2.3.3 Process Control Program The Process Control Program (PCP) contains the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of radioactive waste, based on demonstrated processing of actual or simulated wet or liquid wastes, will be accomplished in such a way as to assure compliance with 10CFR20, 10CFR61, 10CFR71, federal and state regulations, and other requirements governing the transport and disposal of radioactive waste. 3.0 Safety Limits Limits for discharge of gaseous and liquid waste from the Station, and the monitoring of these effluents, are in accordance with Technical Specifications. 4.0 Tests and Inspections 4.1 Waste Process Systems NMP Unit 1 UFSAR Section XII XII-12 Rev. 25, October 2017 The waste processing systems are used on a routine basis and do not require specific testing to assure operability. The effectiveness of design is ultimately demonstrated by the effluent monitors and the environmental monitoring program. 4.2 Filters The exhaust ventilation filters are replaced when the pressure drop across the filters exceeds the normal operating range. Test connections are available for checking the efficiency of newly installed filters. Adequate tests to determine filter efficiency are conducted in accordance with the Technical Specifications. 4.3 Effluent Monitors The effluent monitors will be calibrated periodically to assure that they are accurately detecting effluent activity. 4.3.1 Offgas and Stack Monitors An isotopic analysis is made of a representative sample of gaseous activity downstream of the steam jet air ejectors (SJAE) and at the stack sample point in accordance with the Technical Specifications and the Offsite Dose Calculation Manual (ODCM). These waste gas effluent monitors are calibrated and tested in accordance with the ODCM. 4.3.2 Liquid Waste Effluent Monitor The liquid waste effluent monitor is calibrated* and tested* in accordance with the ODCM. Accounting of liquid waste discharge will be by laboratory analysis and volume measurement as described in the ODCM. B. RADIATION PROTECTION 1.0 Primary and Secondary Shielding 1.1 Design Bases

• Required prior to removal of blank flange in discharge line and until blank flange is replaced.

NMP Unit 1 UFSAR Section XII XII-13 Rev. 25, October 2017 Normal operating conditions determine the major portion of the Station shielding requirements. The exceptions are the main control room, where shielding is determined by the loss-of-coolant accident (LOCA); and the shutdown cooling system, where shielding is determined by shutdown conditions. Shielding is also determined by whether post-accident radiation fields unduly limit personnel access to areas which would: 1) require occupancy to permit an Operator to aid in the mitigation of or recovery from an accident, or 2) unduly degrade the proper operation of safety equipment. Operating condition source configurations are based on activation product activities and neutron fluxes associated with 1779 MWt operation and fission product activities associated with a 1 curie/sec stack release of noble gases (after nominal 30-min holdup). These source values are combined with required occupancy time for each area to determine shielding requirements that would result in personnel exposures less than those specified in 10CFR20. No significant changes in the anticipated doses resulted when Station operation was extended to 1850 MWt. The required occupancy times are arranged into six groups, as shown in Table XII-6. Accident condition source configurations are based on the LOCA accompanied by complete core meltdown, 5-percent drywell leakage per day and no stack release. Shielding calculations were made at 1 hr and at 3 days to evaluate both the short-lived and longer-lived components of the resulting source. Activities used were: Noble Gases Halogens At 1 Hr 3.78 x 105 Ci 4.44 x 104 Ci At 3 Days 1.27 x 107 Ci 3.58 x 105 Ci These total activities were converted into an inventory of noble gases and halogens typical of those expected in the fuel. For shielding computation, the gamma sources were broken down into seven energy groups to separate the important emitters (see Table XII-7). In most cases, the N-16 activation product (with a 6.1 Mev gamma energy) determined gamma shielding requirements. However, in equipment where a significant decay of 7.3-sec N-16 occurs or in which fission products collect, other nuclides contributed to shielding requirements. NMP Unit 1 UFSAR Section XII XII-14 Rev. 25, October 2017 1.2 Design 1.2.1 Reactor Shield Wall The structural design inside the drywell incorporates a hollow cylinder of concrete 2 ft thick around the reactor vessel. The portion of the cylinder at the elevation of the reactor core is heavy concrete, while the remaining portions (which are keyed into the heavy concrete to reduce streaming through the joints) are ordinary concrete. The inside and outside surfaces of the concrete are covered with steel plate, and reinforcing steel is used in the concrete to give structural strength. The cylinder is supported on the same structural concrete as the reactor vessel. This shield is cooled on both surfaces by air circulating from the drywell cooling system. The reactor shield wall acts as a thermal shield during operation to protect the biological shield outside the drywell from thermal damage. During shutdown this shield also acts as a gamma shield around the reactor core during work in the drywell. 1.2.2 Biological Shield The biological shield (located around the exterior of the drywell) varies in thickness from 4 ft to 7 ft of concrete for shielding purposes (it is thicker than 7 ft in some areas for structural reasons). This shield is designed to limit the radiation level from the reactor core and equipment in the drywell to 5 mrem/hr in accessible areas of the reactor building during full-power operation. The biological shield uses bedrock as its main support and is structurally designed to handle the loads of floors, equipment and the higher elevations of the shield itself. Reinforcing steel is included in the design for structural integrity under all credible conditions. Penetrations in the biological shield are designed so that they are not aimed directly at the core or major items of equipment in the drywell. In addition, they are either terminated in shielded cubicles or are shielded with steel flanges to reduce radiation levels in accessible areas to 5 mrem/hr. 1.2.3 Miscellaneous Shielding requirements for the various systems in the Station are given in the sections describing each system. Standard NMP Unit 1 UFSAR Section XII XII-15 Rev. 25, October 2017 concrete is used for the major portion of the shielding. In some locations, structural requirements exceeded shielding requirements and thus determined concrete thicknesses. 1.3 Evaluation Plans for inspection and testing of the shielding are given in Section XII-B.4.1. Since shielding incorporated in the design for accident conditions cannot be tested, the following is a brief description of this shielding design. The activities calculated for the LOCA (Section XII-B.1.1) were spread evenly throughout the available reactor building volume and were treated as two spherical sources (one, with half the activity, for that portion of the reactor building above the operating floor and the other, with the other half of the activity, for that portion below this floor). These spherical sources were then used to evaluate the shielding for the diesel generator room, the battery room and the main control room using typical shielding computational techniques. Structural concrete between the reactor building and both the diesel generator and battery rooms is sufficient to reduce the radiation levels in these areas to less than 5 mrem/hr; therefore, no additional shielding was required for the accident condition. The evaluation of the main control room showed that an 8-in thick concrete shield on the roof and 1-ft thick concrete walls on the north and west sides of the control room were required to ensure that the radiation level in this room is less than 5 mrem/hr at 1 hr and less than 2.5 mrem/hr at 3 days after an accident. (The design radiation level in the control room during normal operation is less than 0.2 mrem/hr.) The main steam line (MSL) break accident and the refueling accident were also evaluated but proved to require less shielding than the LOCA. In the case of the MSL break, the "source" disperses by exfiltration from the turbine building. In the refueling accident, lower activities are released to the reactor building. 2.0 Area Radioactivity Monitoring Systems 2.1 Area Radiation Monitoring System 2.1.1 Design Bases NMP Unit 1 UFSAR Section XII XII-16 Rev. 25, October 2017 1. Personnel in or entering frequently occupied areas must be warned of any significant increase in the general area radiation level. 2. Monitors in areas where radiation levels are not subject to sudden changes give an alarm in the control room only*. The Station intercommunication system will be used to advise personnel of these alarms. 3. Monitors in areas where radiation levels are subject to sudden changes give an alarm both in the control room and in the area where the monitor is located. This alarm must make personnel in the area and those who might enter the area aware of abnormal radiation levels. 2.1.2 Design Monitor Location Forty-nine monitoring units are used for area radiation monitoring in the Station. The locations and ranges of these monitors are given in Table XII-8. Monitor Design A basic General Electric Company (GE) Log Radiation Monitor is used for each location, as described in Table XII-8. The monitor has a four-decade logarithmic readout of radiation level. To cover the various radiation ranges anticipated in the Station, three different models of the basic instrument are used: 0.01 to 102 mrem/hr 0.1 to 103 mrem/hr 10 to 106 mrem/hr (special five-decade monitor) The basic monitor uses a Geiger-Mueller (G-M) detector instrumented to measure dose rate at low radiation levels by the usual pulse counting technique. At higher dose rates, a current generated by the detector is added to the pulse counting circuit output current, thus compensating for the loss of counts due to resolving time losses. The monitor also includes a circuit

• The ARM in the fresh fuel storage vault falls into this category. The 6.6" x 11" center-to-center spacing of bundles in the racks maintains an adequate criticality margin (Keff <0.95).

NMP Unit 1 UFSAR Section XII XII-17 Rev. 25, October 2017 which keeps the instrument indicating full scale for any radiation level greater than full scale. Each monitor has an upscale electronic trip point to indicate an increase in radiation level, and a downscale electronic trip point to indicate instrument malfunction. Both upscale and downscale alarms are annunciated in the control room. Upscale alarms are also annunciated at the locations equipped with local alarms. An indicating meter for each monitor is located in the control room. In addition, each monitor is logged at regular intervals by the Station data logger. In addition to the above area radiation monitors (ARM), the Eberline Radiation Monitoring System (RMS II) detects and measures gamma radiation fields at 17 locations (see Table XII-8 for locations) throughout the radwaste solidification and storage building (RSSB). The dose rate at these locations is displayed locally near the detector and on panel J in the control room. The wall-mounted detector elements use G-M tubes and cover the range from 0.1 to 10,000 mR/hr. Check sources are installed in each detector assembly, which may be operated from the control room panel. Audible and visual alarms are located at each of the 17 locations in the RSSB and in the control room. Calibration check is performed only over the range of interest, namely 10-1000 mR/hr. Accident Analysis Most of the monitors would be of little or no use in the analysis of conditions after an accident since they are designed to monitor the relatively low radiation levels expected during Station operation. However, two of the monitors are designed for use during or following an accident: 1. The monitor in the control room is designed primarily to keep personnel advised of the radiation levels in this area following an accident. 2. The higher-ranged monitor on the fuel pool refueling platform is designed to monitor higher radiation levels that could set the lower-ranged instrument off scale during work in the spent fuel pool. This high-range instrument could also be used in evaluating radiation levels in the reactor building following an accident. 2.1.3 Evaluation NMP Unit 1 UFSAR Section XII XII-18 Rev. 25, October 2017 General Station Operation Even though the area radiation monitors are in areas of relatively low radiation levels, they do not replace portable instrument surveys but rather supplement them. In addition, the monitors are well scattered throughout the Station, so that one monitor does not serve as a "backup" for another monitor. Therefore, there are no requirements for any definite number of monitors to be in service for Station operation. Any malfunctioning monitors will be repaired as soon as is reasonably achievable. Spent Fuel Storage Pool Operation During work in the fuel storage pool in which recently irradiated fuel is handled, the high-range monitor on the refueling platform must be operating to ensure initiation of emergency ventilation system in the event of high radiation. Movement of Monitors If any of the locations chosen prove to be ineffective after a period of Station operation, monitors can be moved to more effective locations. Monitor Range Changes Accidental or unauthorized range changes are precluded because several internal circuit modifications are required to make such changes. Alarm Points Alarm points are set as specified by the General Supervisor Radiation Protection. 2.2 Area Air Contamination Monitoring System 2.2.1 Design Bases 1. Personnel occupying or entering one of the three major buildings (reactor, turbine or waste disposal) should be warned of significant airborne contamination levels. NMP Unit 1 UFSAR Section XII XII-19 Rev. 25, October 2017 2. Monitors actuate an alarm in the control room and at the monitor. The Station intercommunication system will also be used to advise any personnel in the area affected. 2.2.2 Design Monitor Location Constant air monitors are installed in low background areas sampling each of the main exhaust ducts of the three major buildings. Monitor readout and the alarm signal are transmitted to the control room. Additional portable constant air monitors with local readout are provided. Monitor Design Commercially available beta-gamma removable filter constant air monitors are used for the air contamination monitoring system. The monitors have a high-level trip point to indicate an abnormal airborne contamination level and a low-level trip point to indicate instrument malfunction. 2.2.3 Evaluation General Station Operation Since the constant air monitoring system is supplemented by constant air samples (samples taken continuously, with the activities analyzed at the end of the sampling period) and by short-term "grab" samples, it is not essential for Station operation. However, because the constant air monitoring system does provide useful information in a convenient-to-use form, any malfunctioning monitors will be repaired as soon as is reasonable. Movement of Portable Monitors Monitors will be moved in response to operational or maintenance activities to more effective locations, as required. Monitor Alarm Point Changes The alarm trip points are changed as required by varying background radiation levels and airborne activity nuclide analysis. NMP Unit 1 UFSAR Section XII XII-20 Rev. 25, October 2017 3.0 Radiation Protection Procedures for personnel radiation protection are prepared consistent with the requirements of 10CFR20 and are approved, maintained, and adhered to for all operations involving personnel radiation exposure. The following terms are used to designate the various areas within the site boundary: Controlled Area: As defined in 10CFR20, "...means an area, outside of a restricted area but inside the site boundary, access to which can be limited by the licensee for any reason." At Nine Mile Point Nuclear Station - Unit 1 (Unit 1), the Controlled Area includes the area between the site boundary and the security protected area fence. Restricted Area: As defined in 10CFR20, "...means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials..." At Unit 1, the Restricted Area includes the area within the security protected area fence. Radiologically-Controlled Area (RCA): As defined at Unit 1, means major plant areas, access to which is limited for the purpose of protecting personnel from exposure to radiation and contamination. The RCA includes the reactor, turbine, old and new radwaste and offgas buildings, and radwaste solidification and storage building (RSSB). Other RCAs may be established with protective requirements specified by radiation protection supervision in accordance with approved Station procedures. 3.1 Facilities The radiation protection facility and related facilities are shown on Figures III-3 and III-4. 3.1.1 Laboratory, Counting Room and Calibration Facilities A laboratory-counting room complex is used for chemistry and radiation protection analytical work. This complex (located in the turbine building) consists of: NMP Unit 1 UFSAR Section XII XII-21 Rev. 25, October 2017 1. An air lock adjacent to the high-level laboratory, where high-level samples may be delivered. 2. A high-level laboratory, where highly-radioactive samples are analyzed or diluted. 3. A low-level laboratory, where the major portion of Station samples are analyzed and prepared for counting. 4. A counting room, where the activity of the samples is determined. An auxiliary counting room is also provided and currently houses a panoramic irradiator for calibration of dosimetry devices and testing of radiation detection instruments. 5. A technician's office. 6. A laboratory supply storage area for storage of case quantities of chemicals and glassware. The laboratories are equipped with constant air flow hoods so that analyses can be performed with a minimal risk of personnel contamination or ingestion. The hoods are exhausted to the stack and the laboratory sinks drained to the waste neutralizer tank in the waste disposal building to control all laboratory effluents. The remainder of this general building area is used for other radiation protection work. 7. The portable instrument storage area is the central storage point for radiation protection instruments. This room is located in the administration building near the main access point so that personnel can conveniently pick up and return instruments on their normal route into and out of the plant. This room is also used for radiation protection sample analysis using count rate and gamma spectroscopy instruments. 8. The instrument calibration room is equipped with a variety of sources for the calibration of the portable radiation detecting instruments. Instrument calibration is centered around two source wells, 21 ft deep. Co-60 sources are remotely moved up and down the wells to produce the desired radiation level at the top of the well. In addition, a high-level calibration device is provided. NMP Unit 1 UFSAR Section XII XII-22 Rev. 25, October 2017 9. An auxiliary calibration laboratory is provided for calibrating count-rate instruments. 3.1.2 Change Room and Shower Facilities The main change areas (locker rooms) are located in the administration building, adjacent to the entry into the turbine building (Figure III-4). The locker rooms are used by Station personnel to change from street clothes to working clothes, if desired. The shower facility is in the RCA. Protective clothing is processed by contracted service. 3.1.3 Personnel Decontamination Facility The personnel decontamination facility area is located adjacent to the shower facility to enable contaminated personnel to enter from the restricted side, be decontaminated, and then be monitored prior to release. The area is equipped with a sink, shower and appropriate monitoring equipment to assure that no contamination is carried from the RCA. Additional decontamination facilities may be provided, as required. 3.1.4 Tool and Equipment Decontamination Facility 1. Tools and equipment are decontaminated in areas specifically designed for such work, including the turbine building (Figure III-4). In cases where decontamination is relatively easy or where equipment cannot be conveniently transported to this facility, decontamination is done in place. There are also several specialized decontamination areas in the Station to supplement the tool and equipment decontamination facility. a. The control rod drive (CRD) decontamination trough is located outside the drywell personnel lock. b. The irradiated fuel shipping cask is decontaminated in a decontamination pan on the reactor building operating floor. The decontamination pan will not be used for the transfer cask employed during dry cask storage operations. The process area for the transfer NMP Unit 1 UFSAR Section XII XII-23 Rev. 25, October 2017 cask will be located in the reactor building hoist well on elevation 261 ft. c. Drums of waste are decontaminated in the waste disposal building. 2. The tool and equipment decontamination facility is made up of: a. The small tool decontamination area for decontaminating hand tools and small equipment. b. The tool storage room for storing tools used in the RCAs of the Station. c. The equipment decontamination area for decontaminating tools and equipment too large for the small tool decontamination area. It is designed so that equipment is brought in one end, decontaminated and then taken out the other end, thus avoiding recontamination. 3.2 Radiation Control Personnel radiation exposure is kept as low as reasonably achievable (ALARA) by judicious use of shielding, area decontamination and by access control. 3.2.1 Shielding As discussed in Section XII-B.1.0, the shielding is designed to keep personnel exposures to a practical minimum based on expected occupancy times. As Station operation progresses, any shielding deficiencies will be evaluated, and additions to existing shielding will be made where necessary. In any areas where radiation levels are found to be high but shielding cannot be added because of physical layout or operating limitations, personnel will be protected by isolating the area and limiting access, or by decontamination of the area or system. 3.2.2 Access Control Access to the fenced area is controlled by locked doors or gates. Access within this controlled area is further restricted on the basis of radiation levels and the presence of radioactive material. NMP Unit 1 UFSAR Section XII XII-24 Rev. 25, October 2017 Any area in which radioactive material is stored, handled or processed, or in which radiation levels are in excess of those designated in 10CFR20, is designated as a Radiologically-Controlled Area (RCA). In general, all the Station buildings other than the control room area and the administration building are designated as a RCA. Entrances to these areas are normally posted: CAUTION RADIOLOGICALLY-CONTROLLED AREA RADIOACTIVE MATERIALS RWP AND DOSIMETRY REQUIRED FOR ENTRY The requirement for a radiation work permit (RWP) or dosimetry is included on the sign when radiological conditions warrant it. Within the RCA, access to areas of higher radiation is further controlled. 1. Radiation Area Any area in which the radiation level is greater than 5 mrem/hr but less than 100 mrem/hr is designated as a Radiation Area. Personnel exposures in radiation areas are kept to a minimum by use of administrative procedures based on accumulated doses and by keeping time spent in radiation areas as short as possible. Radiation areas may be isolated with yellow and magenta rope and are posted with signs: CAUTION RADIATION AREA 2. High Radiation Area Any area in which the radiation level is greater than 100 mrem/hr is designated as a high radiation area. Entrances to high radiation areas (100 to 1000 mrem/hr) to which personnel require frequent access are barricaded and normally kept locked and may be equipped with an alarm system which will warn the person entering. Entrances to high radiation areas above 1000 mrem/hr shall be provided with locked doors to prevent unauthorized entry, and the hard keys or NMP Unit 1 UFSAR Section XII XII-25 Rev. 25, October 2017 access provided by magnetic keycard shall be maintained under the administrative control of the Shift Manager (SM) or designate on duty and/or the General Supervisor Radiation Protection or designate, and issued to personnel with the appropriate radiation work permit (RWP). Access to high radiation areas that may be reached only by ladders or climbing structures, such as loft spaces above false ceilings or the upper volumes of high rooms, is not controlled by automatic alarms or special barricades. All high radiation areas will be posted with signs: CAUTION OR DANGER HIGH RADIATION AREA Radiation protection personnel make routine surveys of all the accessible areas in the Station to keep abreast of any changes in the radiation levels in these areas. 3.3 Contamination Control Contamination control is achieved in general by physical separation of the contaminated area. 3.3.1 Facility Contamination Control Contamination of the general Station areas is prevented by using the "step-off-pad" technique when leaving areas that are contaminated. Monitoring devices are placed near the step-off-pad so that personnel can check to assure they are not inadvertently carrying some contamination with them. Unless specifically exempted by radiation protection supervision, personnel will monitor themselves prior to each exit from the RCA to assure that no contamination is being carried from the RCA. For maintenance jobs involving high levels of contamination, the installation of plastic or paper on the floor around the equipment to be maintained will permit quick and easy cleanup after the work is completed. Thus, spread of contamination to other equipment or other floor areas is prevented. Radiation protection personnel make routine surveys of the contamination levels in all the accessible areas of the Station to keep abreast of any changes in contamination status. Any areas found contaminated to undesirable levels will be roped off NMP Unit 1 UFSAR Section XII XII-26 Rev. 25, October 2017 and posted. These areas are decontaminated as soon as is reasonable. 3.3.2 Personnel Contamination Control Contamination of personnel is controlled in two ways. First, contamination is prevented from getting into areas where personnel can unknowingly come in contact with it by using the methods described in Section 3.3.1. Second, personnel who enter contaminated areas are protected with special protective clothing. The following types of protective clothing are used: 1. Coveralls - Worn for most work in contaminated areas. 2. Plastic suits - Worn in areas where potential exists for liquid contamination of personnel. 3. Gloves - Cotton gloves are worn for protection against dry contamination and rubber gloves for protection against dry or wet forms of contamination. 4. Shoecovers - Cloth covers are worn for protection against dry contamination; plastic shoecovers for dry or moist contamination; rubber overshoes for dry, moist or wet contamination. 5. Head protection - Caps are worn for protection against low-level contamination; cloth hoods for protection against high-level contamination; plastic hoods for protection against very high or moist contamination. If contamination levels are moderate to high, the various pieces of clothing worn are taped together to prevent contamination from entering the joints. In some cases, double layers of clothing are worn to give additional protection. Normally, most of the Station is accessible to personnel in street clothes or nonradioactive work clothes. To minimize the area in which special protective clothing is worn, such clothing is donned at the job site where it is required. Thus, temporary change areas are set up for special maintenance jobs, or a more permanent change area is established for special areas routinely requiring protective clothing. 3.3.3 Airborne Contamination Control NMP Unit 1 UFSAR Section XII XII-27 Rev. 25, October 2017 Airborne contamination is minimized by keeping floor contamination levels low, and by reducing leaks as much as possible. However, when airborne contamination levels exceed, or if there is potential for exceeding, the values given in 10CFR20, an evaluation of internal dose commitments may be compared to projected whole body exposures and application of respirators based on this evaluation. Allowances are made for the use of respiratory protective equipment in determining whether individuals are exposed to concentrations in excess of the values specified in 10CFR20. The protection factors used do not exceed those authorized by 10CFR20. To assure that these protection factors are provided, the following administrative controls are incorporated in the respiratory protection program. 1. Each respirator user is advised that he may leave the high airborne area for either physical or psychological relief from respirator use, and that he must leave the area in the case of respirator malfunction or any other condition that might cause reduction in the protection afforded the user. 2. Sufficient air samples and other surveys are made to identify the hazards, evaluate individual exposure, and to permit proper selection of respiratory protective equipment. 3. Procedures are established to: a. Assure proper training for the correct use of the various types of respiratory equipment. b. Assure proper maintenance so that the full effectiveness of the respiratory equipment is maintained. (Includes: cleaning, survey, inspection, repair, sanitizing and storage.) 4. Bioassays and/or whole body counts are made on individuals to evaluate individual exposures, and to assess the adequacy of the respiratory protection program. 3.4 Personnel Dose Determinations NMP Unit 1 UFSAR Section XII XII-28 Rev. 25, October 2017 3.4.1 Radiation Dose Monitoring of personnel is accomplished by the use of thermoluminescence dosimeters (TLD), direct-reading dosimeters, electronic dosimeters, and neutron TLD badges. Personnel entering the RCA are issued TLDs and self-reading dosimeters in accordance with 10CFR20 and Station procedures. The TLD readings are normally used as the official record. The TLDs are National Voluntary Laboratory Accreditation Program (NVLAP) certified for gamma and beta radiation and able to differentiate between penetrating and nonpenetrating radiations. TLDs are processed quarterly or in accordance with Station procedures. If a TLD should be lost or damaged, an individual's exposure is estimated using appropriate methods and documented. Direct-reading dosimeters or electronic dosimeters are worn by plant personnel for a day-to-day or job-to-job estimate of personnel exposure. Direct-reading dosimeters are calibrated at a frequency specified in Station procedures, or when damage is suspected. Station personnel are kept advised of the accumulated dose by Radiation Exposure Reports issued at least weekly. Appropriate records of employee doses are maintained at the Station. The internal deposition of radioactive material in personnel working in any of the RCAs of the Station is evaluated at intervals dependent upon their occupational group. This evaluation is made by whole body gamma counting. If a significant deposit is detected, the associated dose will be added to the individual's radiation dose records. Whole body counts will be supplemented by bioassay data when appropriate. 3.5 Radiation Protection Instrumentation 3.5.1 Counting Room Instrumentation The counting room instrumentation includes: 1. Germanium (Ge) detectors. 2. Alpha and beta detectors. 3. G-M type counters with thin window detectors. NMP Unit 1 UFSAR Section XII XII-29 Rev. 25, October 2017 3.5.2 Portable Radiation Instrumentation The portable radiation instruments which are normally stored in the portable radiation instrument storage area of the calibration facility include: 1. Low and intermediate range ion chamber instruments. 2. Portable G-M type instruments. 3. An alpha scintillation instrument. 4. High-range dose rate instruments. 5. A neutron dose rate instrument. 6. An R-meter and a set of standardized ionization thimbles, or equivalent instrument. 7. Small article monitors with plastic scintillation detectors. 3.5.3 Air Sampling Instrumentation The portable air sampling instruments include: 1. Low-volume air samplers equipped to use filter paper, charcoal cartridges, or silver zeolite cartridges. 2. High-volume air samplers equipped to use filter paper, charcoal cartridges, or silver zeolite cartridges. 3.5.4 Personnel Monitoring Instruments The personnel monitoring instruments include: 1. Self-reading dosimeters with a range of 0-200 mrem. 2. Self-reading dosimeters with a range of 0-500 mrem. 3. Count rate meters with G-M detectors. 4. Automatic whole body beta and gamma sensitive contamination monitors. 5. Electronic dosimeters. NMP Unit 1 UFSAR Section XII XII-30 Rev. 25, October 2017 3.5.5 Emergency Instrumentation Instruments are kept in special locations for use as designated by the emergency procedures in event of an emergency where Station instruments are unobtainable. These instruments receive periodic surveillance so that they are known to be functioning correctly. These instruments include: 1. Low-range dose rate meters. 2. High-range dose rate meters. 3. Portable G-M type instruments. 4. Battery-operated high volume air samplers. 5. Self-reading dosimeters with a range of zero to 5 x 103 mrem and zero to 5 x 104 mrem. Procedures and associated training for the accurate determination of airborne iodine concentration in areas within the plant where plant personnel may be present during an accident have been established and implemented, and are maintained to meet or exceed the requirements and recommendations of Section 2.1.8.c of NUREG-0578 (Item III.D.3.3 of NUREG-0737). 4.0 Tests and Inspections 4.1 Shielding Visual inspections of Station shielding were conducted during the construction phase. Their value, however, is limited to locating major defects because of the massive nature of the shielding. During reactor operation, radiation surveys are performed at various power levels. The purpose of these surveys is to assure that: 1. There are no defects or inadequacies in the shielding that might affect personnel exposures during operation of the Station at the same power level as the test. 2. There are no serious defects which might create untenable radiation levels at higher power levels. NMP Unit 1 UFSAR Section XII XII-31 Rev. 25, October 2017 3. Areas in the Station are correctly posted and barricaded as Radiation and High Radiation Areas. These surveys consisted of both gamma and neutron monitoring with appropriate portable instrumentation. Gamma surveys were performed on all Station shielding, while neutron surveys were conducted around the biological shield and associated penetrations. After the survey, key locations were selected for routine radiation surveys throughout the life of the plant. These surveys, while primarily designed to detect changes in radiation levels due to process conditions, also monitor aging effects on shielding integrity. 4.2 Area Radiation Monitors Each area radiation monitor is tested to: 1. Determine that the monitor is correctly wired into the control room. 2. Calibrate the monitor so that the control room readout instrumentation indicates true radiation levels. (For the GE monitors, radiation sources are placed at reproducible geometries on each monitor detector to set the calibration of at least two points on the four-decade scale). 3. Set upscale and downscale alarm trip points. 4. Determine that both the control room and the local alarm (when so equipped) function correctly. Steps 2, 3 and 4 are repeated periodically to assure that calibration and alarm setpoints are correct. 4.3 Area Air Contamination Monitors Each area air contamination monitor is tested to: 1. Determine that the monitor is correctly wired into the control room. 2. Calibrate the monitor so that meter readings can be interpreted in terms of c/cc. (Filter papers impregnated with known quantities of appropriate NMP Unit 1 UFSAR Section XII XII-32 Rev. 25, October 2017 radionuclides are placed on the detector section of each monitor to set the calibration at no less than two points over the range of the monitor.) 3. Set upscale/downscale (GE) and high/alert (Eberline) alarm trip points. 4. Determine that both control room and the monitor alarms function correctly. Steps 2, 3 and 4 are repeated periodically to assure that calibration and alarm setpoints are correct. 4.4 Radiation Protection Facilities 4.4.1 Ventilation Air Flows Ventilation air flows in the radiation protection facilities are checked as part of the turbine building ventilation tests. 4.4.2 Instrument Calibration Well Shielding The instrument calibration well shielding was tested when the sources were installed. Station surveys of nearby areas ensure continued shielding integrity. 4.5 Radiation Protection Instrumentation The following instrumentation is tested and calibrated at a frequency specified in Station procedures, with deviations, not exceeding annually, allowed based on documented instrument reliability: 1. Counting room instrumentation. 2. Portable radiation instruments. 3. Personnel monitoring instruments (except self-reading dosimeters). 4. Emergency instruments. 5. Air samplers. 6. Self-reading dosimeters. NMP Unit 1 UFSAR Section XII XII-33 Rev. 25, October 2017 Tests and calibration include (where applicable): 1. Calibration with appropriate calibrated radioactive sources. 2. Calibration of air flow rates with a flow rate measuring system. NMP Unit 1 UFSAR Section XII XII-34 Rev. 25, October 2017 TABLE XII-1 FLOWS AND ACTIVITIES OF MAJOR SOURCES OF GASEOUS ACTIVITY Design Flow Stack Activity Source (cfm) Radionuclides at 1850 MWt Offgas 22 Xenon - 133 Activities from System Krypton - 85m this source will Krypton - 88 vary depending on plant operating conditions and will not exceed the limits specified in 10CFR20. Steam 1,900 Same as above Activities from Packing this source will vary depending on plant operating conditions and will not exceed the limits specified in 10CFR20. Building 180,000 Xenon - 138 Activities from Exhausts plus those this source will above vary depending on plant operating conditions and will not exceed the limits specified in 10CFR20. NMP Unit 1 UFSAR Section XII XII-35 Rev. 25, October 2017 TABLE XII-2 QUANTITIES AND ACTIVITIES OF LIQUID RADIOACTIVE WASTES Liquid Waste (Gal/Day) Activity Level (µCi/ml) Maximum Normal Maximum Normal Low-Conductivity Liquid Wastes Drywell Recirculation Pump Seals Valves Maintenance Reactor Building Pump Seal Leakage (Cleanup, Rod Drive, Sludge) Valves (Scram, Feed, etc.) Resin Transfer and Filter Drains Maintenance Turbine Building Feed Pump Seals Condensate Pump Seals Sample Drains Maintenance Waste Disposal Building Offgas Drains Resin Transfers --- --- 4,500 --- --- --- 600 --- --- --- 2,100 --- 62,000 5,400 1,730 0 990 1,980 1,500 0 720 720 720 0 5 12,500 2 2 2 2x10-2 2x10-1 2x10-5 1 2x10-5 2x10-3 2x10-4 --- 100 --- 10-1 10-1 0 10-3 10-2 10-6 10-2 10-6 10-4 10-5 10-4 --- 2x10-2 High-Conductivity Liquid Wastes Drywell Floor Drains Control Rod Drive Drains Reactor Building Floor Drains Maintenance Drains Turbine Building Floor Drains Decontamination Laboratory Drains Shower Facility --- --- --- --- --- --- --- 300 500 2,000 2,300 Varies 2,500 0 500 50 2x10-5 2x10-5 10-2 --- 10-5 --- --- --- 10-6 10-6 10-4 10-3 10-6 1 10-4 <10-5 NMP Unit 1 UFSAR Section XII XII-36 Rev. 25, October 2017 TABLE XII-2 (Cont'd.) Liquid Waste (Gal/Day) Activity Level (µCi/ml) Maximum Normal Maximum Normal High-Conductivity Liquid Wastes (cont'd.) Waste Disposal Building Floor Drains Resin Cleaning and Backwashing Solutions Decontamination Filter Backwash --- 1,200 --- 1,324 500 696 100 883 --- 4x10-1 --- 3.45x10-1 10-4 2x10-3 1 4.28x10-1 NOTE: In some cases, maximum quantities are not listed because the exact values are unknown. These maximum values are expected to be close to the values listed as normal. NMP Unit 1 UFSAR Section XII XII-37 Rev. 25, October 2017 TABLE XII-3 ANNUAL SOLID WASTE ACCUMULATION AND ACTIVITY Approximate Shipment Accumulation Activity Filter Sludge Normal Volume 327 cu ft/yr 933 curies Spent Resins Condensate (outage) 500 cu ft/yr 10 curies Condensate (nonoutage) 270 cu ft/yr 10 curies Cleanup 400 cu ft/yr 120 curies Concentrated Waste Normal Volume 2,000 gal/yr 14 curies Dry Wastes Compressible (outage) 20,560 cu ft/yr 0.124 curies Compressible (nonoutage) 5,000 cu ft/yr 0.124 curies Noncompressible 2,500 cu ft/yr 0.025 curies Filter septa (prior to incineration) 125.7 cu ft/yr *

• Filter septa shipment activity will be determined based on plant experience after the new CFS is in service.

NMP Unit 1 UFSAR Section XII XII-38 Rev. 25, October 2017 TABLE XII-4 LIQUID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS Component Qty Capacity Concentrated Waste Tank 1 8,000 gal Chemical Addition Tank 1 600 gal Drywell Equipment Sump 2 2,000 gal* Drywell Floor Drain Sump 1 2,000 gal Electric Boiler 1 16,000 lb/hr Filter Aid Tank 1 470 gal Floor Drain Collector Tank 1 10,000 gal Floor Drain Filter 1 300 gpm Floor Drain Sample Tank 2 20,000 gal* Precoat Tank 1 560 gal Reactor Building Equipment Drain Tank 1 5,000 gal Reactor Building Floor Drain Sump 6 24,600 gal* Turbine Building Equipment Drain Tank 2 5,900 gal* Turbine Building Floor Drain Sump 8 8,200 gal* Utility Collector Tank 1 16,000 gal Waste Building Equipment Drain Sump 1 2,300 gal Waste Building Floor Drain Sump 3 3,200 gal Waste Collector Tank 1 25,000 gal Waste Collector Filter 1 300 gpm Waste Concentrator 1 20 gpm Waste Demineralizer 1 2,300 gal Waste Neutralizer Tank 1 15,000 gal Waste Sample Tank 2 50,000 gal* Waste Surge Tank 1 70,000 gal

• Total capacity.

NMP Unit 1 UFSAR Section XII XII-39 Rev. 25, October 2017 TABLE XII-5 SOLID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS Component Qty Capacity Filter Sludge Storage Tank 3 23,000 gal* Spent Resin Tank 1 4,000 gal

• Total capacity.

NMP Unit 1 UFSAR Section XII XII-40 Rev. 25, October 2017 TABLE XII-6 OCCUPANCY TIMES Expected Maximum Design Radiation Zone Occupancy Level Designation (hrs/wk) (mrem/hr) 1 Controlled Restricted >100 Accessibility 2 Controlled Infrequent 100 Accessibility 3 1-5 20 4 20 5 5 50-67 1.5 6 Continuous Occupancy 0.2 NMP Unit 1 UFSAR Section XII XII-41 Rev. 25, October 2017 TABLE XII-7 GAMMA ENERGY GROUPS Gamma Energy Representative Limits Energy for Group Group (Mev) (Mev) I 0.1 - 0.4 0.4 II 0.4 - 0.9 0.8 III 0.9 - 1.35 1.3 IV 1.35 - 1.8 1.7 V 1.8 - 2.2 2.2 VI 2.2 - 2.6 2.5 VII >2.6 Energy of Nuclide NMP Unit 1 UFSAR Section XII XII-42 Rev. 25, October 2017 TABLE XII-8 AREA RADIATION MONITOR DETECTOR LOCATIONS Range of Monitor Bldg/ Monitor No. Location Elev (mR/hr) 1 SE Plant Entrance TB 261 0.01-100 2 New Fuel Room RB 318 0.01-100 3 Control Room Admin Bldg AB 277 0.01-100 4 In-Plant I&C Shop TB 277 0.01-100 5 Generator Area TB 300 W 0.1-1000 6 Shaft Pump Area TB 300 E 0.1-1000 7 Cond Pump Vlvs Condenser Bay TB 261 NE 0.1-1000 8 Outside MSIV Room TB 261 0.1-1000 9 N of Battery Board Rooms TB 261 0.1-1000 10 Cond Demin Valve Room TB 257 0.1-1000 11 Regen Room TB 261 0.1-1000 12 Truck Bay TB 261 0.1-1000 13 Deleted 13A Condensate Filter System TB 300 0.1-1000 14 Old RW Bldg S of Stairs EL 229 0.1-1000 15 Old RW Bldg Control Room EL 261 0.1-1000 16 Old RW Bldg Door to Pusher Rm EL 261 0.1-1000 17 Inner TIP Room RB 249 0.1-1000 NMP Unit 1 UFSAR Section XII XII-43 Rev. 25, October 2017 TABLE XII-8 (Cont'd.) Range of Monitor Bldg/ Monitor No. Location Elev (mR/hr) 18 West End of Shield Wall RB 340 0.1-1000 19 RX Bldg - NE Corner EL 198 0.1-1000 20 Closed Loop Cooling Area RB 298 0.1-1000 21 Cleanup Pump Area RB 261 0.1-1000 22 RX Bldg - NE EL 281 0.1-1000 23 CRD Accumulator Area RB 237 0.1-1000 24 Large Equip Decon Room TB 261 0.1-1000 25 RX Bldg - East Wall EL 340 0.1-1000 26 High Level Chem Lab TB 261 0.1-1000 27 RX Bldg - NW EL 318 0.1-1000 28 North Instr Room RB 237 0.1-1000 29 Refuel Bridge (Low Range) RB 340 0.1-1000 - Refuel Bridge (High Range) 10-106 (Process Mon) 30 New RW Bldg N of Decon Pan EL 261 0.1-1000 31 New RW Bldg West Wall EL 247 0.1-1000 32 New RW Bldg South Wall EL 229 0.1-1000 33 Offgas Bldg W of Stairs EL 229 0.1-1000 RSSB 1 Cement Fill Area RSSB 244 0.1-10,000 NMP Unit 1 UFSAR Section XII XII-44 Rev. 25, October 2017 TABLE XII-8 (Cont'd.) Range of Monitor Bldg/ Monitor No. Location Elev (mR/hr) 2 Valve & Pump Room West RSSB 244 0.1-10,000 3 Valve & Pump Room East RSSB 244 0.1-10,000 4 Electric Switchgear Room RSSB 244 0.1-10,000 5 Feed Equipment Area Volume Red Sys - South RSSB 261 0.1-10,000 6 Feed Equipment Area Volume Red Sys - North RSSB 261 0.1-10,000 7 Access Way - South RSSB 261 0.1-10,000 8 Access Way - North RSSB 261 0.1-10,000 9 North-South Truck Bay RSSB 261 0.1-10,000 10 East-West Truck Bay RSSB 261 0.1-10,000 11 HVAC Supply Fan - South RSSB 281 0.1-10,000 12 HVAC Recirc. Atmos. - West RSSB 281 0.1-10,000 13 HVAC Exhaust Fans - North RSSB 281 0.1-10,000 14 Concentrated Waste Tank Access - West RSSB 281 0.1-10,000 15 Concentrated Waste Flush Tank Access - East RSSB 281 0.1-10,000 16 HVAC Recirc. Atmos. Cleanup System - South RSSB 292 0.1-10,000 17 HVAC Exhaust System Char. Filter Area RSSB 292 0.1-10,000 SPENT RESIN TANK FIGURE XII-1 20 UFSAR Revi2007 October U.S. NUCLEAR REGULATORY COMMISSION DOCKET 50-220 LICENSE DPR-63 NINE MILE POINT NUCLEAR STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT (UPDATED) OCTOBER 2017 REVISION 25 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section XIII XIII-i Rev. 25, October 2017 SECTION XIII CONDUCT OF OPERATIONS XIII-1 A. ORGANIZATION AND RESPONSIBILITY XIII-1 1.0 Offsite Organization XIII-1 1.1 Station Organization XIII-1 1.1.1 This Section Deleted XIII-1 1.1.2 This Section Deleted XIII-1 1.1.3 This Section Deleted XIII-1 2.0 Nine Mile Point Nuclear Station, LLC, Organization XIII-2 2.1 This Section Deleted XIII-2 2.2 This Section Deleted XIII-2 2.3 This Section Deleted XIII-2 3.0 This Section Deleted XIII-9 4.0 Operating Shift Crews XIII-9 5.0 Qualifications of Staff Personnel XIII-9 B. QUALIFICATIONS AND TRAINING OF PERSONNEL XIII-10 1.0 This Section Deleted XIII-10 2.0 This Section Deleted XIII-10 3.0 This Section Deleted XIII-10 4.0 Training of Personnel XIII-10 4.1 General Responsibility XIII-10 4.2 This Section Deleted XIII-10 4.3 This Section Deleted XIII-10 4.3.1 This Section Deleted XIII-10 4.3.2 This Section Deleted XIII-10 4.3.3 This Section Deleted XIII-10 4.3.4 This Section Deleted XIII-10 4.3.5 This Section Deleted XIII-10 4.3.6 This Section Deleted XIII-10 4.3.7 This Section Deleted XIII-10 4.3.8 For Operations Director and Operations Shift Superintendent XIII-10 4.4 Training of Licensed Operator Candidates/Licensed NRC Operator Retraining XIII-11 5.0 Cooperative Training With Local, State and Federal Officials XIII-12 C. OPERATING PROCEDURES XIII-13 D. EMERGENCY PLAN AND PROCEDURES XIII-14 NMP Unit 1 UFSAR Section Title Section XIII XIII-ii Rev. 25, October 2017 E. SECURITY XIII-15 F. RECORDS XIII-16 1.0 Operations XIII-16 1.1 Control Room Log XIII-16 1.2 Shift Manager's Log XIII-16 1.3 Radwaste Log XIII-16 1.4 Waste Quantity Level Shipped XIII-16 2.0 Maintenance XIII-17 3.0 Radiation Protection XIII-17 3.1 Personnel Exposure XIII-17 3.2 By-Product Material as Required by 10CFR30 XIII-17 3.3 Meter Calibrations XIII-17 3.4 Station Radiological Conditions in Accessible Areas XIII-17 3.5 Administration of the Radiation Protection Program and Procedures XIII-17 4.0 Chemistry and Radiochemistry XIII-17 5.0 Special Nuclear Materials XIII-18 6.0 Calibration of Instruments XIII-18 7.0 Administrative Records and Reports XIII-18 G. REVIEW AND AUDIT OF OPERATIONS XIII-19 1.0 This Section Deleted XIII-19 1.1 This Section Deleted XIII-19 2.0 This Section Deleted XIII-19 2.1 This Section Deleted XIII-19 3.0 This Section Deleted XIII-19 NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section XIII XIII-iii Rev. 25, October 2017 XIII-1 Deleted XIII-2 Deleted NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section XIII XIII-iv Rev. 25, October 2017 XIII-1 thru 5 DELETED (Replaced by the Quality Assurance Topical Report, NO-AA-10) NMP Unit 1 UFSAR Section XIII XIII-1 Rev. 25, October 2017 SECTION XIII CONDUCT OF OPERATIONS A. ORGANIZATION AND RESPONSIBILITY Exelon Generation Company, LLC, is a limited liability and is responsible for the safe, reliable, and efficient operation of its nucler facilities. In addition, Exelon is responsible for appropriate standards, programs, processes, management controls, and support for the nuclear facilities. In keeping with these responsibilities, Exelon is committed to providing sufficient personnel having appropriate qualifications to both operate and technically support the facility. 1.0 Offsite Organization The Exelon corporate organization and its functions and responsibilities are described in Chapter 1 of the Quality Assurance Topical Report, NO-AA-10, as revised. 1.1 Station Organization The Station organization is as described in NO-AA-10. 1.1.1 Deleted 1.1.2 Deleted 1.1.3 Deleted 2.0 Nine Mile Point Nuclear Station, LLC, Organization This section describes the structure, function, and responsibilities of the onsite organizations established to operate and maintain the plant. The onsite and offsite independent review committees are described in NO-AA-10. Unit 1 and Unit 2 operations are independent of each other, including backshift operation. Only licensed individuals may direct licensed activities. The lines of authority are described in administrative procedures. 2.1 Deleted NMP Unit 1 UFSAR Section XIII XIII-2 Rev. 25, October 2017 2.2 Deleted 2.3 Deleted 3.0 Deleted 4.0 Operating Shift Crews The Operations Department is under the direction of the Director Operations who reports to the Plant Manager. The Director Operations has the following direct reports: the Unit Shift Operation Superintendent, the Reactor Engineering Manager, and the Senior Manager Operations Support and Services. The Unit Shift Operations Superintendent is directly responsible for supervision of plant operations including management oversight of shift operations. The Unit Shift Managers report directly to the Unit Shift Operations Superintendent. The Director Operations, Unit Shift Operations Superintendents, Senior Manager Operations Support and Services, Operations Support Manager, or Operation Services Manager must possess a Senior Reactor Operator (SRO) license. Station operating shift complements are described in station procedures. The following additional requirements apply: 1. At least one licensed Operator shall be in the control room when fuel is in the reactor. During reactor operation, this licensed Operator shall be present at the controls of the facility. 2. A licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling shall be responsible for all movement of new and irradiated fuel within the site boundary. 5.0 Qualifications of Staff Personnel Each member of the unit staff, with the exception of the Operator license applicants and the Radiation Protection Manager, shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions. The Radiation Protection Manager shall meet or exceed the qualifications of RG 1.8, September 1975. The licensed operators shall comply only with the requirements of 10 CFR 55. A retraining and replacement training program for the facility staff shall be maintained under the direction of the Manager Training, and shall meet or exceed the recommendations and NMP Unit 1 UFSAR Section XIII XIII-3 Rev. 25, October 2017 requirements of Section 5.5 of ANSI N18.1-1971 and of 10CFR55, and shall include familiarization with relevant industry operational experience. B. QUALIFICATIONS AND TRAINING OF PERSONNEL 1.0 (This section deleted) 2.0 (This section deleted) 3.0 (This section deleted) 4.0 Training of Personnel 4.1 General Responsibility The training program includes development and conduct of technician, operator, and support programs. Additional training and educational programs are presented as needed. Training programs are prepared to include formal objectives and written lesson plans. During development of programs, close liaison is maintained with appropriate line managers to ensure that content, test material, mode of presentation, and schedule are appropriate. As additional training needs are identified, new training programs or lessons are developed to meet requirements not covered by existing programs. Program responsibilities are described in NO-AA-10. 4.2 Deleted 4.3 Deleted 4.3.1 Deleted 4.3.2 Deleted 4.3.4 Deleted 4.3.5 Deleted 4.3.6 Deleted 4.3.7 Deleted 4.3.8 For Operations Director and Operations Shift Superintendent As a minimum, either the Operations Director or the Operations Shift Superintendent shall hold a SRO license. The Operations Director, who in lieu of meeting the SRO license requirements of NMP Unit 1 UFSAR Section XIII XIII-4 Rev. 25, October 2017 ANSI N18.1-1971, shall: 1) hold a SRO license at the time of appointment, or 2) have held a SRO license at Unit 1 or at a similar unit, or 3) have been certified for equivalent SRO knowledge. 4.4 Training of Licensed Operator Candidates/Licensed NRC Operator Retraining Detailed training programs for Unit 1 Operations are designed to provide initial training, requalification training, and continuing training at all levels of the Operations organization. These programs fulfill the requirements included in the following documents: 10CFR50, Licensing of Production and Utilization Facilities 10CFR55, Operators Licenses NUREG-0737, Clarification of TMI Action Plan Requirements NUREG-1021, Operator Licensing Examiner Standards ANSI N18.1-1971, Selection and Training of Nuclear Power Plant Personnel ANSI/ANS 3.4-1983, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants The training program is designed in accordance with accreditation programs described in the latest approved ACAD recommendations, and uses a simulation facility acceptable to the NRC under 10CFR55. Entry into these training programs is controlled by the Operations organization. Eligibility criteria for license candidates are contained in station procedures. 5.0 Cooperative Training With Local, State and Federal Officials A detailed Site Emergency Plan and Procedures for Nine Mile Point Nuclear Station has been submitted to the NRC. Included in this document are the training procedures involving local, state and federal officials. C. OPERATING PROCEDURES NMP Unit 1 UFSAR Section XIII XIII-5 Rev. 25, October 2017 The Station operating staff has prepared written operating procedures to be used for all normal operating conditions. Changes may be initiated by the Station operating staff subject to approval by the Operations supervisory staff, and by the department manager for the functional area of the procedure, or higher levels of management as governed by administrative procedures. Procedures cover operation of major systems such as starting the entire Station from "cold" conditions. Other procedures cover less extensive systems in detail. Still another type of procedure instructs the Operator in the methods of operating individual pieces of equipment, such as regeneration of resin in a demineralizer. The format for all operating procedures is essentially the same. Each procedure is prefaced with the technical limitations of the system or equipment, as set forth in the Technical Specifications, OL, or 10CFR20. Other data helpful to operation, such as a system description and plant operating requirements, are included in a separate section of the procedure. Details of operation are then set down in a stepwise procedure. Prior to startup, prepared lists are checked off by the Operators. These vary in degree with the extent of the period preceding the startup. The emergency operating procedures (EOP) and the severe accident procedures (SAP) have been developed, validated and implemented in accordance with the requirements of NUREG-0737 Supplement 1. The EOPs/SAPs are prepared using the guidance provided by the ccident Guidelines (BWROG EPG/SAG). The EOPs/SAPs are symptom oriented rather than event based. They address conditions beyond the design basis. They provide guidance for the entire range of available systems. This "defense-in-depth" approach provides for safe shutdown of the plant in all postulated events, including anticipated transients without scram (ATWS), thus preventing or mitigating the consequences of any accident or malfunction. In the event of an unlikely, yet credible, accident situation which might involve radioactivity release to the public domain, the Shift Manager, in accordance with written procedures, is responsible for notifying Station supervision and outside authorities. It is recognized that a program of this type must be rehearsed; therefore, planned nuclear incident drills are held periodically to review established procedures, personnel assignments and relations with outside authorities. D. EMERGENCY PLAN AND PROCEDURES NMP Unit 1 UFSAR Section XIII XIII-6 Rev. 25, October 2017 The prime objectives of emergency planning are to: (1) Develop a plan and implementing procedures that will provide the means for mitigating the consequences of emergencies (including very low probability events) in order to protect the health and safety of the general public and site personnel and to prevent damage to property, and (2) Ensure operational readiness of emergency preparedness capabilities. The Nine Mile Point Nuclear Station Emergency Plan assures that all emergency situations, including those which involve radiation or radioactive material are handled logically and efficiently. It covers the entire spectrum of emergencies from minor, localized emergencies to major emergencies involving action by offsite emergency response agencies and organizations. The Emergency Plan Implementing Document provides a single source of pertinent and significant information and data and the procedures that would be required by or useful for various emergency response agencies and organizations in the event of an emergency. In Implementing document consolidates and integrates specific material detailed in documents such as the Nine Mile Point Nuclear Station Emergency Plan, the State Plans, and the Various County Plans. This Emergency Plan has been developed in accordance with the provisions of 10 CFR 50, Appendix E, and 10 CFR 50.47. Other guidance and sources of information used in the development of the Emergency Plan have been identified in the Exelon Nuclear Standardized Radiological Emergency Plan Annex for Nine Mile Point. The Site Emergency Plan and Implementing Procedures have been submitted to the NRC under separate cover. Changes to the Site Emergency Plan are made in accordance with the requirements of 10CFR50.54(q). E. SECURITY A detailed Nine Mile Point Nuclear Station Physical Security, Safeguards Contingency, and Security Training and Qualification Plan, identified as safeguards information and withheld from public disclosure in accordance with 10CFR73.21, has been submitted to the NRC. The security plan described above details the measures taken to provide adequate Site and Station security and conforms to NMP Unit 1 UFSAR Section XIII XIII-7 Rev. 25, October 2017 10CFR73.55. Changes to the security plan are made in accordance with the requirements of 10CFR50.54(p) or 10CFR50.90, as applicable. F. RECORDS 1.0 Operations The following logs will be maintained by the operating staff as a part of the Station records. When electronic logs are used, the control room log and the SM log may be combined. 1.1 Control Room Log Shall contain all information pertaining to changing core reactivity during all modes of reactor operation, including rod manipulation, orifice modifications, control rod testing, etc. Also, entries affecting Station outputs, changes in auxiliary equipment, unusual condition, line trips, annunciator signals not recorded on data logger, etc., will be entered in this log. The log shall contain the date and time of all entries and the name of the Chief Shift Operator (CSO) or other authorized personnel only. The control room log is to be treated as a legal document subject to being entered in a court record. All entries in this log shall be by the Operator on duty or his Supervisor. No other entries are authorized. Included with the control room log is a fuel log in which specific detailed fuel moves, channel changes, and in-core instrumentation changes are recorded. 1.2 Shift Manager's Log Shall contain an overall summary of Station operation including the name of the SM on duty, the Operators and Auxiliary Operators on duty, major equipment not in service or inoperable, and the date and time of all entries. Also note any Operator surveillance tests run and deviations from acceptance criteria. The log may be written by the Control Room Supervisor (CRS) or a CRS/SM in training, or other designee, but must be signed and acknowledged by the SM. 1.3 Radwaste Log The log shall contain pertinent information associated with the radwaste facility operation. 1.4 Waste Quantity Level Shipped NMP Unit 1 UFSAR Section XIII XIII-8 Rev. 25, October 2017 Solid waste and resins removed from site. 2.0 Maintenance The Maintenance Director will be responsible for maintaining a record of maintenance performed on all pertinent equipment in a maintenance log. 3.0 Radiation Protection Radiation Protection Manager will be responsible for the following records. 3.1 Personnel Exposure 1. Dosimeter readings, daily 2. Thermoluminescence dosimeter (TLD) record, quarterly 3. Continuous exposure record conforming to 10CFR20 4. Appropriate records and forms required in 10CFR20 3.2 By-Product Material as Required by 10CFR30 3.3 Meter Calibrations of all survey meters, environmental monitors and monitors affecting radioactive discharge. 3.4 Station Radiological Conditions in Accessible Areas 1. Radiation levels 2. Contamination levels 3. Airborne activity 3.5 Administration of the Radiation Protection Program and Procedures 4.0 Chemistry and Radiochemistry The Chemistry Manager is responsible for primary and secondary system Chemistry and Radiochemistry including monitoring and control of liquid and gaseous radiological effluents. 5.0 Special Nuclear Materials NMP Unit 1 UFSAR Section XIII XIII-9 Rev. 25, October 2017 The special nuclear materials records will be maintained and reported in conformity with 10CFR70. 6.0 Calibration of Instruments The calibration of instruments and controls, both nuclear and conventional, will be recorded, as well as maintenance performed on them. 7.0 Administrative Records and Reports 1. Investigations of abnormal operation will be prepared in report form and distributed to interested parties. 2. Records will be kept of all changes to equipment or procedures. 3. Reports of production and pertinent operating data with a summary of items of interest will be produced at regular intervals and distributed to interested parties and to those who audit Station operations. 4. Reports of exposure to individuals, loss or theft of licensed material, etc., as outlined in 10CFR20 will be reported in the time and manner specified. G. REVIEW AND AUDIT OF OPERATIONS The functions, composition and responsibilities of those organizations responsible for performing the nuclear safety review and audit of Nine Mile Point Nuclear Station are delineated in the Quality Assurance Topical Report, NO-AA-10, as revised. 1.0 Deleted 1.1 Deleted 2.0 Deleted 2.1 Deleted 3.0 Deleted Refer to the Quality Assurance Topical Report, NO-AA-10, as revised for the qualifications of essential managerial positions NMP Unit 1 UFSAR Section XIII XIII-10 Rev. 25, October 2017 and to the applicable Human Resources procedures for comparable ANSI/ANS 18.1-1971 positions for individuals responsible for programs and systems that ensure the safe and successful operation of the facility. Changes to these documents are evaluated in accordance with the applicable change control process. NMP Unit 1 UFSAR Section XIII XIII-11 Rev. 25, October 2017 TABLE XIII-2 This Table has been deleted. Station operating shift complements are described in station procedures.

NMP Unit 1 UFSAR Section XIV XIV-i Rev. 25, October 2017 TABLE OF CONTENTS Section Title SECTION XIV INITIAL TESTING AND OPERATIONS A. TESTS PRIOR TO INITIAL REACTOR FUELING B. INITIAL CRITICALITY AND POSTCRITICALITY TESTS 1.0 Initial Fuel Loading and Near-Zero Power Tests at Atmospheric Pressure 1.1 General Requirements 1.2 General Procedures 1.3 Core Loading and Critical Test Program 2.0 Heatup from Ambient to Rated Temperature 2.1 General 2.2 Tests Conducted 3.0 From Zero to 100 Percent Initial Reactor Rating 4.0 Full-Power Demonstration Run 5.0 Comparison of Base Conditions 6.0 Additional Tests at Design Rating 7.0 Startup Report NMP Unit 1 UFSAR Section XIV XIV-1 Rev. 25, October 2017 SECTION XIV INITIAL TESTING AND OPERATIONS A. TESTS PRIOR TO INITIAL REACTOR FUELING A program of preoperational testing of equipment and systems was carried out prior to initial operation of the Station. The purpose of this program was to demonstrate that construction was in accordance with specifications. The test program included those checks, adjustments, calibrations, and operations required to assure that fuel loading and Station operation could be undertaken safely and efficiently. The following paragraphs outline the scope of testing prior to initial reactor fueling. Process instruments including remote transmitters, recorders and indicating instruments were checked, tested and calibrated. Control systems, both automatic and manual, were tested for proper installation and operation. Power distribution systems and equipment, both normal and emergency ac and dc batteries, were tested for electrical continuity and proper operation. It was demonstrated that Station auxiliaries can be automatically transferred from normal to reserve supply. The emergency diesel generator units were placed in operation manually, operated at full load and also tested for automatic starting. The reactor vessel, the reactor shutdown cooling system and the reactor cleanup system were filled with demineralized water and the systems operated to check out pump, valve and equipment operability. The main coolant water was heated to a minimum of the nil ductility reference temperature (RTNDT) plus 60°F for the initial hydro test. Piping and support hangers were checked while thermal expansion progressed. Control and instrumentation systems were rechecked during operation of these systems. The liquid poison system was filled with demineralized water and operated to check pumps, heater, and valves and verify injection time of the liquid poison. The tank was filled with the required concentration of sodium pentaborate before fuel loading. The core spray system was operated using demineralized water to demonstrate component and system performance in both manual and NMP Unit 1 UFSAR Section XIV XIV-2 Rev. 25, October 2017 automatic modes. The required response time of the system to initiating signals was verified. The inlet and outlet valves on the emergency cooling system were checked for proper operation and their response to simulated initiating signals was verified. In latter phases of the startup program, the emergency cooling system inlet valves were opened from a hot pressurized condition to verify their operability in this mode. The reactor safety valves were preset and tested before installation. The solenoid-actuated relief valves were tested for proper operation after installation and their response to proper initiating signals was verified. The Station protective system was functionally tested to assure that it was operating properly. All sensors were calibrated and adjusted with a test signal to verify that, at the proper signal level, a reactor scram signal and/or other protective actions would be initiated. As a signal level corresponding to a scram condition or protective action was reached, scram initiation, protective action and annunciator indications were verified. The neutron monitoring instrumentation was thoroughly checked for proper installation and operation. Test signals were used for initial calibration adequate for initial fuel loading and near zero power testing. A simulated test signal was supplied to the input of each neutron flux amplifier to check the interlock and scram function of these instruments. The chamber retraction mechanisms for source and intermediate range instrumentation were tested to demonstrate that the mechanisms operated satisfactorily both electrically and mechanically. The traversing in-core probe (TIP) system mechanisms were test operated to demonstrate that the detector probes could be inserted and withdrawn to each assigned local power range monitor (LPRM) in-core monitor string, and that the associated control and position readout equipment was operating properly. The complete control rod hydraulic system, including drives, scram valves, dump volume, piping, regulating valves and interlocks, was checked for proper operation. Adjustments were made to provide desired operating speed of drive mechanisms. Measurements of scram times were made on all drives. Additional adjustments and testing of the system were also performed during initial pressurization. NMP Unit 1 UFSAR Section XIV XIV-3 Rev. 25, October 2017 The radioactive waste disposal system was tested to assure that all pumps, valves, filters, demineralizers and associated controls and instrumentation operated in accordance with design specifications. Remote handling devices were checked for ease of operation and maintenance before the actual handling of radioactive materials was required. The reactor recirculation system was tested, including motor generator, pump motor, pumps and valves. The control system for the pumps was thoroughly tested and adjusted for proper operation. Additional testing of the recirculation system was performed during low-power and full-power testing. The fuel storage, head cavity and core internal storage areas, and the associated piping were checked for leakage. Dummy fuel assemblies were run through a complete cycle from the new fuel storage area into the reactor core, shuffled within the core, and finally removed to the spent fuel storage pool. The fuel pool filtering and cooling system was checked to verify proper operation of pumps, filter system and cooling system. All Station service systems were checked and operated for conformance to specifications, performance, cleanliness and accessibility of valves, controls, instruments and specialties. These service systems included the following: 1. Reactor building closed loop cooling water (RBCLCW) system. 2. Turbine building closed loop cooling water (TBCLCW) system. 3. Service water (SWP) system. 4. Makeup water (MWS) system. 5. Breathing, instrument and service air system. 6. Fire protection system. The turbine generator and its auxiliaries (condensate and feedwater system and condenser circulating water system) could not be fully tested under actual service conditions because of lack of a large steam source. However, reasonable effort was made to assure their proper functioning. For example, all instruments were calibrated and the control systems, both NMP Unit 1 UFSAR Section XIV XIV-4 Rev. 25, October 2017 automatic and manual, were tested for proper installation and operation insofar as possible. The Station intercommunication system and the area alarm system were checked for proper operation. The overall installation of the system was reviewed to determine if all areas of the Station were adequately covered and, if required, modifications were made. The drywell and pressure suppression chamber were tested for leakage as further described in Section VI-F. As described in this section, all penetrations, vacuum relief valves and isolation valves were tested for leakage. The containment isolation valves were tested to assure proper closure times, and their response to initiating signals was verified. The containment spray system was operated at rated flow conditions to demonstrate component and system performance in both manual and automatic modes. The sprays were operated initially with water and subsequently with air so as to establish a point of reference for comparison with future periodic air tests. The drywell cooling system and the containment inerting system were checked to prove out the controls and to demonstrate their ability to operate satisfactorily. The reactor building normal and emergency ventilation systems were tested and adjusted and automatic initiation of the emergency ventilation system was verified. The reactor building was isolated and it was demonstrated that in-leakage would not exceed allowable values under the partial vacuum drawn by the emergency ventilation system fans. Tests of the emergency ventilation particulate and charcoal filters were made to verify power gasketing to assure minimization of bypass leakage around the filters. The heating and ventilation systems for all buildings and the control room were checked and adjusted for proper operation. Radiation monitoring instrumentation, both process and area monitors, were checked for proper installation and operation. Test signals and calibration sources were used for calibration of instruments and adjustment of setpoints for scram and alarm signals. The control room readout instrumentation was checked and annunciator indications were verified. NMP Unit 1 UFSAR Section XIV XIV-5 Rev. 25, October 2017 Tests of systems and components not specifically listed above were performed as required to assure that all systems were complete and operable and that the Station was ready for initial fuel loading. B. INITIAL CRITICALITY AND POSTCRITICALITY TESTS 1.0 Initial Fuel Loading and Near-Zero Power Tests at Atmospheric Pressure 1.1 General Requirements Accurate knowledge of reactor parameters and characteristics, as required for safe operation of the Station, were determined by an extensive program of tests and measurements executed before, during and after fuel was initially loaded into the reactor. The fuel, control curtains and control rods were subjected to extensive quality control tests and measurements during manufacturing to assure that physical properties were those specified. The nuclear characteristics of these components were calculated by methods which were compared with results of experiments in the Vallecitos Atomic Laboratory's critical facilities, including measurements of similar or identical components. In addition, startup, test and operating data from boiling water reactors (BWR) in commercial operation, and other measurements throughout the nuclear industry, were used to confirm the applicability of the analytical methods. 1.2 General Procedures Detailed step-by-step written procedures were prepared for conducting the startup program. Special attention was given to the quality and completeness of the knowledge about the core prior to and during the tests, with emphasis on nuclear safety. The following general procedures were directed toward performing the startup program as safely and efficiently as was practical. 1. The preoperational functional testing of all necessary equipment had to be completed and the equipment had to be in safe and operable condition. 2. The control rods and curtains were all installed in the core prior to fuel loading. 3. The vessel was filled with demineralized water to a minimum of 8 ft above the top of the core. Within 24 NMP Unit 1 UFSAR Section XIV XIV-6 Rev. 25, October 2017 hr prior to loading the reactor, the water was analyzed for dissolved poisons with special attention to boron. 4. The startup test source was installed prior to the first fuel loading, in a water gap where it was surrounded by fuel by the end of the first loading step. 5. The reactor protection system (RPS) was operating before fuel loading started. The minimum requirements were: a. At least three low-level detectors, B-10s or fission chambers, were in waterproof containers in the vessel. These detectors were to monitor the neutron flux at levels below the response level of the permanently installed RPS detectors. These temporary chambers were in the reactor protection circuit. A one-out-of-three scram logic was employed. b. A demonstrated signal, principally neutron dependent, was available from each detector. c. To the extent practical, detectors were placed to detect fission neutrons rather than source neutrons. 6. A communications link between the loading crew and the Operator in the control room was installed and operable. 7. Identical tag boards were available for use at the loading area and in the control room for logging the location of all readily movable core components (e.g., source, temporary neutron detectors, fuel assemblies by number, etc.) at all times. 8. At every stage during the initial loading and in the fully loaded configuration, the control rod system, augmented by curtains, provided a shutdown margin. This requirement was demonstrated by ascertaining that the reactor was subcritical with the strongest worth control rod fully withdrawn. NMP Unit 1 UFSAR Section XIV XIV-7 Rev. 25, October 2017 9. Loading steps consisted of four assemblies per step until a core size slightly greater than four times the calculated minimum critical size was achieved. 10. The standard fuel loading technique was separated into the method for intermediate size cores (less than four times the minimum critical size), and the method for larger cores (greater than four times minimum critical size). The sequence that was followed using an intermediate size core was: a. Select the control rod cell (up to four fuel assemblies) to be loaded. b. Fully withdraw and then insert the nearest strong control rod which was surrounded by fuel. This demonstrated that the core was subcritical with a strong worth control rod withdrawn. c. Functionally test the control rod drive (CRD) in the cell to be loaded to ascertain integrity and assure coupling. d. Load the selected control rod cell. e. Again withdraw and insert the control rod used in step b. This was to demonstrate that the core was still subcritical with a strong worth control rod withdrawn, and give assurance that no gross loading errors were made. If the core was ever found to be supercritical in this step, it would have been shut down promptly and the last loading increment removed. f. Fully withdraw the control rod from the cell loaded in step d to functionally test and demonstrate both the proper mechanical operation of the control rod and at least a one-stuck-rod shutdown margin. If the core was ever found to be supercritical in this step, it would have been shut down promptly and the last loading increment removed. g. The fuel loading in a large core follows the same sequence as described above. The difference is that up to four control rod cells, one cell per core quadrant (a total of 16 fuel assemblies), NMP Unit 1 UFSAR Section XIV XIV-8 Rev. 25, October 2017 may be loaded each step. Quadrant symmetry would be created and maintained, if practical, to assure maximum decoupling between the control cells being loaded, and to permit loading the full core with a minimum number of steps. 1.3 Core Loading and Critical Test Program The initial fuel loading and critical testing was performed near zero power, at atmospheric pressure, with the vessel open. The tests that were performed during this phase of the startup program included the following: 1. Chemical and Radiochemical tests were conducted to determine and establish proper water conditions prior to initial operation and to maintain these throughout the test program. Base or background radioactivity levels were determined at this time for use in fuel element failure detection and long range activity buildup. 2. CRD System tests were performed on all drives prior to fuel loading to assure proper operability and to measure and adjust operating speeds. Drive line friction and scram times were determined for all drives at zero reactor pressure. Functional testing of each drive was performed just previous to and following the fuel loading in each cell to ascertain drive and coupling integrity. 3. The RPS System was functionally tested prior to fuel loading to assure all necessary sensors and system components were available and operating properly. 4. Fuel Loading then commenced, following detailed step-by-step written procedures, based on the general procedures in paragraph 1.2. The core was assembled (with curtains) to its full size. 5. Stuck Rod Shutdown Margin Beginning with the intermediate-sized core, it was demonstrated that the reactor was subcritical by more than a specified amount with the strongest worth control rod withdrawn. The stuck rod reactivity margin requirement is a limitation on the amount of reactivity which can be loaded into the core. The magnitude of the margin was chosen with consideration for credible reactivity NMP Unit 1 UFSAR Section XIV XIV-9 Rev. 25, October 2017 changes after the test, and for the accuracy of measurement. The test had three parts: 1) the determination of the strongest worth control rod, 2) the experimental or analytical calibration of an adjacent control rod, and 3) the demonstration of subcriticality with the strongest worth rod fully withdrawn and the second at a position equal to the margin. This demonstration was made for the fully loaded core and with selected smaller core loadings. 6. The specified Startup Control Rod Withdrawal Sequence was evaluated to verify that the stated criteria of safety, simplicity and operating requirements were met during routine cold startups. The reactor was made critical by withdrawing control rods in the specified sequence, and reactivity addition rates were measured near critical. A small number of nonstandard sequences were evaluated. 7. Operating Source and Instrument Adequacy was evaluated. Data taken with the source range monitoring (SRM) instrumentation and the operating sources was compared with stated criteria on noise, signal-to-noise ratio, and response to change in core reactivity. Additional tests of the nuclear instrument adequacy were made during the power phases of the startup program. 8. Initial Power Calibration of Nuclear Instruments was performed to provide a power level calibration for the nuclear instruments that was adequate for this phase of the test program. The waterproofed detectors used in the vessel for fuel loading and startup tests were removed, and the RPS control was transferred to the in-core nuclear instrumentation with the installation of the operating sources. Power calibrations were provided from basic nuclear data. 9. The Rod Worth Minimizer (RWM) was evaluated during tests using standard and nonstandard rod withdrawal patterns, and during simulated rod withdrawal errors, to establish or verify the RWM trip setpoints. Additional tests of the RWM were made during the power phases of the startup program. 2.0 Heatup from Ambient to Rated Temperature NMP Unit 1 UFSAR Section XIV XIV-10 Rev. 25, October 2017 2.1 General Following satisfactory completion of the core loading and low-power test program, the readily movable components in the core were visually verified for proper installation, and the additional in-vessel hardware such as the steam separator and dryer assemblies were installed. The reactor vessel head was then installed. This was followed by a hydrostatic test at less than 1050 psig to assure satisfactory sealing of the vessel head. The drywell head was installed next and shield plugs placed over it. Then a sequence of tests were performed to confirm some of the nuclear steam supply system (NSSS) characteristics as the temperature and pressure were increased. Sufficient tests were performed at each incremental step increase in power or change in pressure, and the tests and operating procedures were evaluated to assure that the succeeding change in operating conditions could be made safely with a minimum of extrapolation. 2.2 Tests Conducted 1. CRD System tests were continued by measuring scram times on four drives at two intermediate pressures, scram times and drive time functional tests on a representative sample of drives at rated reactor pressure, and on four drives without accumulators at rated reactor pressure. 2. The RPS System was functionally tested at rated temperature and pressure to assure all sensor and system components were available and operating properly. 3. The first criticality after head installation was essentially a repetition of one or more of the control rod startup configurations previously used in the critical tests to assure that there had been no change in the conditions. The Instrument Adequacy was evaluated by observing instrument response to power changes during normal power increases. 4. The effectiveness of the RWM was demonstrated during rod withdrawals in the heating power ranges. 5. Recirculation Pump operating characteristics were evaluated to assure stability as the reactor was heated. Various combinations of pumps were operated. NMP Unit 1 UFSAR Section XIV XIV-11 Rev. 25, October 2017 6. Reactor Vessel Temperatures were monitored during heatup and cooldown to assure that temperature differences were acceptable. 7. Initial Thermal Power Calibration of the Nuclear Instruments was determined from the rate of system heating while at constant neutron flux, with the reactor near 5 percent of rated thermal power. 8. The Reactor Power Response to Reactivity Additions was determined by control rod movement at pressures of about 0, 300 and 900 psig. 9. The chosen Rod Withdrawal Sequence was checked to verify that the stated criteria on reactivity addition rates in the heating range were satisfied. 10. System Thermal Expansion checks were made to verify freedom of major equipment and piping to move. 11. Initial Radiation Level Measurements were made to assure that safe and acceptable radiation levels existed in accessible areas. 12. The Emergency Cooling System operational test was performed by manually actuating the emergency cooling system when the Station was operating at a power level slightly in excess of the emergency cooling system capability. A heat balance test for each half of the emergency cooling system was performed. (A similar test was performed upon restart after replacing the emergency condenser tube bundles in 1997.) 13. Operational testing of Reactor Auxiliary Equipment was done to check the functional performance and yield performance data. 3.0 From Zero to 100 Percent Initial Reactor Rating Initial reactor rating was increased to 100 percent in increments of up to 25 percent. The turbine was placed in service during this phase. The test program included the following, but not necessarily at each increment of power. 1. The Instrument Adequacy was evaluated by observing the power tracking accuracy and response of the power NMP Unit 1 UFSAR Section XIV XIV-12 Rev. 25, October 2017 range monitoring instruments (average power range monitor (APRM) and LPRM) during power changes. The SRM and intermediate range monitor (IRM) response to neutron flux and gamma field was evaluated during subsequent planned hot restarts. 2. The RWM was checked for its effectiveness during rod withdrawals as power was increased. 3. The Recirculation Pumps were operated and tripped off in various combinations at various reactor power levels to determine system steady-state and transient characteristics. 4. System Thermal Expansion measurements were continued as various systems were placed into operation at rated pressure. 5. Reactor Auxiliary Systems tests were continued. 6. Radiation Level measurements were continued. 7. The effectiveness of the Steam Separators and Dryers was evaluated at several reactor powers by using carryover analyses of steam samples. No internal vessel measurements were made. 8. The Turbine Startup and Main Bypass System test was performed by bringing the turbine to speed and checking component operation, and by operating the bypass system. Final balancing and adjustments to the turbine generator and bypass controls were made. 9. The Turbine Reheaters and valves were tested. 10. Turbine Trip tests were performed to measure reactor response and main bypass system performance. 11. Generator Trip tests were performed to verify the turbine overspeed trip and the resultant reactor response. 12. System Transient tests determined the effects of various system transients on reactor stability. Such tests as level variation, pressure regulator setpoint change and the like were performed. NMP Unit 1 UFSAR Section XIV XIV-13 Rev. 25, October 2017 13. Radiation Level measurements were continued. 14. Chemical and Radiochemical measurements were continued. 15. Power Calibration of Nuclear Instrumentation was done using conventional heat balance techniques. 16. The LPRM calibration was conducted with neutron flux measurements from moveable chambers (TIP). The LPRM was calibrated in terms of local fuel rod surface heat flux. 17. Core Performance - Gross thermal power, core power distributions from the TIP system, gross recirculation flow, feedwater flow and other system parameters were determined. These experimental data combined with analytical models were used to determine minimum critical heat flux ratio, channel flow, steam quality, maximum heat flux, etc., at various levels up to initial reactor rating. 18. Power Calibration of Control Rods was done to obtain reference relationships between control rod positions and reactor power in standard sequences. 19. Axial Power Distribution measurements were made using the TIP system, before and after significant changes were made in power and rod pattern. 20. Output Variation with flow was demonstrated by making ramp load changes with recirculation flow control. 21. An Operational Rod Pattern change was made to demonstrate a safe change in operating rod pattern at a high Station load. Two alternate rod patterns were specified in order to maintain uniform burnup in the core. 4.0 Full-Power Demonstration Run A demonstration test consisting of a 100-hr run at initial reactor rating to show that design and contractual requirements had been met took place during this phase. Tests that were accomplished during this phase included: NMP Unit 1 UFSAR Section XIV XIV-14 Rev. 25, October 2017 1. Gross Reactor Power Output and Station Net Heat Rate. 2. Radiation Survey. 3. Chemical and Radiochemical Analysis. 4. Axial Power Distribution and LPRM Calibration. 5. Core Performance Evaluation. 5.0 Comparison of Base Conditions A set of initial conditions was established at 1538 MWt before the power increase (to 1850 MWt) was initiated, which served as a basis for comparison with subsequent tests. These base conditions included chemical and radioactivity conditions at typical locations, radiation measurements, APRM calibrations, LPRM response characteristics, axial power distribution measurements and a core performance evaluation. During the power increase program, these tests were repeated at an intermediate point of about 1700 MWt and again at the design 1850 MWt rating. 6.0 Additional Tests at Design Rating 1. The response of the reactor and the turbine governor system to the operating pressure regulator and backup pressure regulator was determined. A step change was made to the operating pressure regulator setpoint and the response of the system was evaluated. The backup pressure regulator was tested by increasing the operating pressure regulator setpoint rapidly until the backup regulator took over control. The response of the system was evaluated and regulator settings were optimized. 2. The ability of the pressure regulator to minimize the reactor pressure disturbance during an external change in the steam flow was determined. One of the turbine bypass valves was tripped open by a test switch. The response of the system was evaluated to aid in making final adjustments to the pressure regulator. 3. The reactor water level was decreased rapidly from a high level and then increased rapidly from a low level. The response of the system was analyzed for stability. NMP Unit 1 UFSAR Section XIV XIV-15 Rev. 25, October 2017 4. A centrally located control rod was moved and signals from nearby LPRM chambers analyzed for reactor stability. 5. Recirculation flow rate was decreased and increased to demonstrate that response to both slow and fast ramp changes in flow rate was stable. 6. The five recirculation pumps were tripped simultaneously from the full-flow conditions and the results analyzed to ensure that reactor thermal limits were not exceeded. 7. The turbine was tripped and the reactor and bypass system responses were evaluated. 7.0 Startup Report A summary report of plant startup and power escalation testing shall be submitted to the NRC in accordance with 10CFR50.4 following 1) receipt of an operating license, 2) amendment to the license involving a planned increase in power level, 3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and 4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. Startup reports shall be submitted within 1) 90 days following completion of the startup test program, 2) 90 days following resumption or commencement of commercial power operation, or 3) 9 months following initial criticality, whichever is earliest. If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

NMP Unit 1 UFSAR LIST OF EFFECTIVE FIGURES SECTION XV Figure Revision Number Number Section XV EF XV-1 Rev. 25, October 2017 XV-1 17 XV-2 14 XV-3 14 XV-4 14 XV-5 14 XV-6 14 XV-7 14 XV-8 14 XV-9 14 XV-10 14 XV-11 14 XV-12 14 XV-13 14 XV-14 14 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section XV XV-i Rev. 25, October 2017 SECTION XV SAFETY ANALYSIS A. INTRODUCTION B. BOUNDARY PROTECTION SYSTEMS 1.0 Transients Considered 2.0 Methods and Assumptions 3.0 Transient Analysis 3.1 Turbine Trip Without Bypass 3.1.1 Objectives 3.1.2 Assumptions and Initial Conditions 3.1.3 Comments 3.1.4 Results 3.2 Loss of 100°F Feedwater Heating 3.2.1 Objective 3.2.2 Assumptions and Initial Conditions 3.2.3 Results 3.3 Feedwater Controller Failure- Maximum Demand 3.3.1 Objective 3.3.2 Assumptions and Initial Conditions 3.3.3 Comments 3.3.4 Results 3.4 Control Rod Withdrawal Error 3.4.1 Objective 3.4.2 Assumptions and Initial Conditions 3.4.3 Comments 3.4.4 Results 3.5 Main Steam Line Isolation Valve Closure (With Scram) 3.5.1 Objective 3.5.2 Assumptions and Initial Conditions 3.5.3 Comments 3.5.4 Results 3.6 Inadvertent Startup of Cold Recirculation Loop 3.6.1 Objective 3.6.2 Assumptions and Initial Conditions 3.6.3 Comment 3.6.4 Results 3.7 Recirculation Pump Trips 3.7.1 Objectives 3.7.2 Assumptions and Initial Conditions 3.7.3 Comments NMP Unit 1 UFSAR Section Title Section XV XV-ii Rev. 25, October 2017 3.7.4 Results 3.8 Recirculation Pump Stall 3.8.1 Objective 3.8.2 Assumptions and Initial Conditions 3.8.3 Comments 3.8.4 Results 3.9 Recirculation Flow Controller Malfunction - Increase Flow 3.9.1 Objective 3.9.2 Assumptions and Initial Conditions 3.9.3 Comments 3.9.4 Results 3.10 Flow Controller Malfunction - Decrease Flow 3.10.1 Objective 3.10.2 Assumptions and Initial Conditions 3.10.3 Comments 3.10.4 Results 3.11 Inadvertent Actuation of One Solenoid Relief Valve 3.11.1 Objectives 3.11.2 Assumptions and Initial Conditions 3.11.3 Comments 3.11.4 Results 3.12 Safety Valve Actuation (Overpressurization Analysis) 3.12.1 Objectives 3.12.2 Assumptions and Initial Conditions 3.12.3 Comments 3.12.4 Results 3.13 Feedwater Controller Malfunction (Zero Demand) 3.13.1 Objective 3.13.2 Assumptions and Initial Conditions 3.13.3 Comments 3.13.4 Results 3.14 Turbine Trip with Partial Bypass (Low Power) 3.14.1 Objectives 3.14.2 Assumptions and Initial Conditions 3.14.3 Comments 3.14.4 Results 3.15 Turbine Trip with Partial Bypass (Full Power) NMP Unit 1 UFSAR Section Title Section XV XV-iii Rev. 25, October 2017 3.15.1 Objectives 3.15.2 Assumptions and Initial Conditions 3.15.3 Comments 3.15.4 Results 3.16 Inadvertent Actuation of One Bypass Valve 3.16.1 Objectives 3.16.2 Assumptions and Initial Conditions 3.16.3 Comments 3.16.4 Results 3.17 One Feedwater Pump Trip and Restart 3.17.1 Objective 3.17.2 Assumptions and Initial Conditions 3.17.3 Comments 3.17.4 Results 3.18 Loss of Main Condenser Vacuum 3.19 Loss of Electrical Load (Generator Trip) 3.19.1 Objectives 3.19.2 Assumptions and Initial Conditions 3.19.3 Comments 3.19.4 Results 3.20 Loss of Auxiliary Power 3.20.1 Objective 3.20.2 Assumptions and Initial Conditions 3.20.3 Comments 3.20.4 Results 3.21 Pressure Regulator Malfunction 3.21.1 Objective 3.21.2 Assumptions and Initial Conditions 3.21.3 Comments 3.21.4 Results 3.22 Instrument Air Failure 3.22.1 Objective 3.22.2 Assumptions and Initial Conditions 3.22.3 Comments 3.22.4 Results 3.23 Dc Power Interruptions 3.23.1 Objective 3.23.2 Assumptions and Initial Conditions 3.23.3 Comments 3.23.4 Results 3.24 Failure of One Diesel Generator to Start NMP Unit 1 UFSAR Section Title Section XV XV-iv Rev. 25, October 2017 3.24.1 Objective 3.24.2 Assumptions and Initial Conditions 3.24.3 Comments 3.24.4 Results 3.25 Power Bus Loss of Voltage 3.25.1 Objective 3.25.2 Assumptions and Initial Conditions 3.25.3 Comments 3.25.4 Results C. STANDBY SAFEGUARDS ANALYSIS 1.0 Main Steam Line Break Outside the Drywell 1.1 Identification of Causes 1.2 Accident Analysis 1.2.1 Valve Closure Initiation 1.2.2 Feedwater Flow 1.2.3 Core Shutdown 1.2.4 Mixture Level 1.2.5 Subcooled Liquid 1.2.6 System Pressure and Steam-Water Mass 1.2.7 Mixture Impact Forces 1.2.8 Core Internal Forces 1.3 Radiological Effects 1.3.1 Radioactivity Releases 1.3.2 Meteorology and Dose Rates 2.0 Loss-of-Coolant Accident 2.1 Introduction 2.2 Input to Analysis 2.2.1 Operational and ECCS Input Parameters 2.2.2 Single Failure Study on ECCS Manually-Controlled Electrically-Operated Valves 2.2.3 Single Failure Basis 2.2.4 Pipe Whip Basis 2.2.5 Degraded Voltage 2.3 This section deleted 2.4 ECCS-LOCA Performane Analysis 2.4.1 Computer Codes 2.4.2 Description of Model Changes 2.4.3 Analysis Procedure 2.4.3.1 Deleted NMP Unit 1 UFSAR Section Title Section XV XV-v Rev. 25, October 2017 2.4.3.2 Deleted 2.4.4 Analysis Results 3.0 Refueling Accident 3.1 Identification of Causes 3.2 Accident Analysis 3.3 Radiological Effects 3.3.1 Fission Product Releases 3.3.2 Meteorology and Dose Rates 4.0 Control Rod Drop Accident 4.1 Identification of Causes 4.2 Accident Analysis 4.3 Designed Safeguards 4.4 Procedural Safeguards 4.5 Radiological Effects 4.5.1 Fission Product Releases 4.5.2 Meteorology and Dose Rates 5.0 Containment Design Basis Accident 5.1 Original Recirculation Line Rupture Analysis - With Core Spray 5.1.1 Purpose 5.1.2 Analysis Method and Assumptions 5.1.3 Core Heat Buildup 5.1.4 Core Spray System 5.1.5 Containment Pressure Immediately Following Blowdown 5.1.6 Containment Spray 5.1.7 Blowdown Effects on Core Components 5.1.8 Radiological Effects 5.1.8.1 Fission Product Releases 5.1.8.2 Meteorology and Dose Rates 5.2 Original Containment Design Basis Accident Analysis - Without Core Spray 5.2.1 Purpose 5.2.2 Core Heatup 5.2.3 Containment Response 5.3 Design Basis Reconstitution Suppression Chamber Heatup Analysis 5.3.1 Introduction 5.3.2 Input to Analysis 5.3.3 DBR Suppression Chamber Heatup Analysis 5.3.3.1 Computer Codes 5.3.3.2 Analysis Methods NMP Unit 1 UFSAR Section Title Section XV XV-vi Rev. 25, October 2017 5.3.3.3 Analysis Results for Containment Spray Design Basis Assumptions 5.3.3.4 Analysis Results for EOP Operation Assumptions 5.3.4 Conclusions 6.0 New Fuel Bundle Loading Error Analysis 6.1 Identification of Causes 6.2 Accident Analysis 6.3 Safety Requirements 7.0 Meteorological Models Used in Accident Analyses 7.1 Introduction 7.2 Atmospheric Dispersion Factor Calculations 7.2.1 Offsite - EAB and LPZ 7.2.2 Control Room and Technical Support Center (Excluding MSLB) 7.2.3 Control Room - MSLB Puff Release 7.3 Summary of Results 7.4 Exfiltration 7.5 Secondary Containment Drawdown 7.5.1 Introduction 7.5.2 Analysis 7.5.3 Results D. REFERENCES NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section XV XV-vii Rev. 25, October 2017 XV-1 TABLE DELETED XV-2 TRIP POINTS FOR PROTECTIVE FUNCTIONS XV-3 thru TABLES DELETED XV-4 XV-5 BLOWDOWN RATES XV-6 REACTOR COOLANT CONCENTRATIONS (µCi/gm) XV-7 TABLE DELETED XV-7a MSLB ACCIDENT ANALYSIS INPUTS AND ASSUMPTIONS XV-7b MSLB ACCIDENT RELEASE RATES XV-8 MAIN STEAM LINE BREAK ACCIDENT DOSES XV-9 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS XV-9a TABLE DELETED XV-9b CORE SPRAY SYSTEM FLOW PERFORMANCE ASSUMED IN THE LOCA ANALYSIS XV-10 ECCS SINGLE VALVE FAILURE ANALYSIS XV-11 SINGLE FAILURES CONSIDERED IN LOCA ANALYSIS XV-12 TRACG-LOCA Licensing Results for Nine Mile Point 1 XV-13 thru TABLES DELETED XV-21 XV-21a ANALYSIS ASSUMPTIONS FOR NINE MILE POINT 1 CALCULATIONS XV-22 ACTIVITY RELEASED TO THE REACTOR BUILDING FOLLOWING THE FHA (CURIES) NMP Unit 1 UFSAR Table Number Title Section XV XV-viii Rev. 25, October 2017 XV-23 UNIFORM UNFILTERED STACK DISCHARGE RATES FROM 0 TO 2 HR AFTER THE FHA (CURIES/SEC) XV-24 FUEL HANDLING ACCIDENT DOSES XV-25 FHA ANALYSIS INPUTS AND ASSUMPTIONS XV-26 CRD ACCIDENT ANALYSIS INPUTS AND ASSUMPTIONS XV-27 CRDA NOBLE GAS RELEASE XV-28 CRDA HALOGEN RELEASE XV-29 CONTROL ROD DROP ACCIDENT DOSES XV-29a WETTING OF FUEL CLADDING BY CORE SPRAY XV-29b POST-LOCA AIRBORNE DRYWELL FISSION PRODUCT INVENTORY (CURIES) XV-29c POST-LOCA REACTOR BUILDING FISSION PRODUCT INVENTORY (CURIES) XV-29d POST-LOCA DISCHARGE RATES (CURIES/SEC) XV-30 CORE FISSION PRODUCT INVENTORY XV-31 LOCA ANALYSIS INPUTS AND ASSUMPTIONS XV-32 LOSS-OF-COOLANT ACCIDENT DOSES XV-32a SIGNIFICANT INPUT PARAMETERS TO THE DBR CONTAINMENT SUPPRESSION CHAMBER HEATUP ANALYSIS XV-33 TABLE DELETED XV-34 TABLE DELETED XV-34a RELEASE/INTAKE ELEVATIONS XV-34b RELEASE/INTAKE DISTANCE AND DIRECTIONS XV-35 TABLE DELETED NMP Unit 1 UFSAR Table Number Title Section XV XV-ix Rev. 25, October 2017 XV-35a X/Q VALUES FOR THE CONTROL ROOM XV-35b X/Q VALUES FOR THE TECHNICAL SUPPORT CENTER XV-35c OFFSITE X/Q VALUES FOR GROUND-LEVEL RELEASES XV-35d OFFSITE X/Q VALUES FOR ELEVATED RELEASES XV-36 REACTOR BUILDING LEAKAGE PATHS NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section XV XV-x Rev. 25, October 2017 XV-1 STATION TRANSIENT DIAGRAM XV-2 FIGURE DELETED XV-3 PLANT RESPONSE TO LOSS OF 100°F FEEDWATER HEATING XV-4 thru FIGURES DELETED XV-7 XV-8 STARTUP OF COLD RECIRCULATION LOOP - PARTIAL POWER XV-9 RECIRCULATION PUMP TRIPS (1 PUMP) XV-10 RECIRCULATION PUMP TRIPS (5 PUMPS) XV-11 RECIRCULATION PUMP STALL XV-12 FLOW CONTROLLER MALFUNCTION (INCREASED FLOW) XV-13 FLOW CONTROLLER MALFUNCTION DECREASING FLOW XV-14 INADVERTENT ACTUATION OF ONE SOLENOID RELIEF VALVE XV-15 thru FIGURES DELETED XV-16 XV-17 FEEDWATER CONTROLLER MALFUNCTION - ZERO FLOW XV-18 TURBINE TRIP WITH PARTIAL BYPASS INTERMEDIATE POWER XV-19 TURBINE TRIP WITH PARTIAL BYPASS XV-20 INADVERTENT ACTUATION OF ONE BYPASS VALVE XV-21 ONE FEEDWATER PUMP TRIP AND RESTART XV-22 LOSS OF ELECTRICAL LOAD XV-23 LOSS OF AUXILIARY POWER NMP Unit 1 UFSAR Figure Number Title Section XV XV-xi Rev. 25, October 2017 XV-24 PRESSURE REGULATOR MALFUNCTION XV-25 MAIN STEAM LINE BREAK - COOLANT LOSS XV-26 thru FIGURES DELETED XV-56 XV-56A thru FIGURES DELETED XV-56C XV-56D LOSS-OF-COOLANT ACCIDENT - WITH CORE SPRAY CLADDING TEMPERATURE XV-56E LOSS-OF-COOLANT ACCIDENT DRYWELL PRESSURE XV-56F LOSS-OF-COOLANT ACCIDENT SUPPRESSION CHAMBER PRESSURE XV-56G LOSS-OF-COOLANT ACCIDENT CONTAINMENT TEMPERATURE - WITH CORE SPRAY XV-57 CONTAINMENT DESIGN BASIS CLAD TEMPERATURE RESPONSE - WITHOUT CORE SPRAY XV-58 CONTAINMENT DESIGN BASIS METAL-WATER REACTION XV-59 CONTAINMENT DESIGN BASIS CLAD PERFORATION WITHOUT CORE SPRAY XV-60 CONTAINMENT DESIGN BASIS CONTAINMENT TEMPERATURE - WITHOUT CORE SPRAY XV-60A DBR ANALYSIS SUPPRESSION POOL AND WETWELL AIRSPACE TEMPERATURE RESPONSE - CONTAINMENT SPRAY DESIGN BASIS ASSUMPTION XV-60B DBR ANALYSIS SUPPRESSION POOL AND WETWELL AIRSPACE TEMPERATURE RESPONSE - EOP OPERATION ASSUMPTIONS XV-61 REACTOR BUILDING MODEL XV-62 EXFILTRATION VS. WIND SPEED - NORTHERLY WIND NMP Unit 1 UFSAR Figure Number Title Section XV XV-xii Rev. 25, October 2017 XV-63 REACTOR BUILDING DIFFERENTIAL PRESSURE XV-64 EXFILTRATION VS. WIND SPEED - SOUTHERLY WIND XV-65 REACTOR BUILDING - ISOMETRIC XV-66 REACTOR BUILDING - CORNER SECTIONS XV-67 REACTOR BUILDING - ROOF SECTIONS XV-68 REACTOR BUILDING - PANEL TO CONCRETE SECTIONS XV-69 REACTOR BUILDING - EXPANSION JOINT SECTIONS XV-70 REACTOR BUILDING EXFILTRATION - NORTHERLY WIND XV-71 REACTOR BUILDING EXFILTRATION - SOUTHERLY WIND XV-72 REACTOR BUILDING DIFFERENTIAL PRESSURE XV-73 REACTOR BUILDING PRESSURE VS. TIME BY REACTOR BUILDING ELEVATION XV-74 REACTOR BUILDING PRESSURE VS. TIME BY REACTOR BUILDING ELEVATION (FOCUSED ON THE INITIAL 2.5 HR) NMP Unit 1 UFSAR Section XV XV-1 Rev. 25, October 2017 SECTION XV SAFETY ANALYSIS A. INTRODUCTION Two types of safety analyses are considered. For the first group discussed in Section XV-B, a complete range of single-failure-caused transients which are abnormal but reasonably expected during the life of the plant were analyzed as part of the original Final Safety Analysis Report (FSAR). Currently, only the limiting transient events are reanalyzed for each reload evaluation. These analyses include those transients which would result in a significant reduction of minimum critical power ratio (MCPR). The transients most likely to limit operation because of MCPR considerations are: 1. Turbine trip without bypass, 2. Loss of feedwater heating, 3. Feedwater controller failure (maximum demand), and 4. Control rod withdrawal error. The main steam isolation valve (MSIV) closure (safety valve actuation overpressurization) event is discussed in Sections V and XV-B.3.12. Also analyzed during each reload is the MSIV closure with scram transient to show compliance with Nuclear Regulatory Commission (NRC) requirements for overpressurization events of moderate frequency. All of the 24 original transients conducted (with applicable revisions) and the loss of feedwater heating transient are presented herein. Section XV-C, Standby Safeguards Analysis, consists of those accidents of a more serious nature leading to a loss of primary system integrity or cladding failure. B. BOUNDARY PROTECTION SYSTEMS 1.0 Transients Considered NMP Unit 1 UFSAR Section XV XV-2 Rev. 25, October 2017 The transients analyzed herein are listed on Figure XV-1, Station Transient Diagram. This illustration is a simplified one-line diagram of the Station divided into three main areas consisting of the reactor system, main steam line-feedwater system and the turbine-generator-condenser-electrical system. In each area, the transients considered are listed and their effect on the reactor designated. As shown on Figure XV-1, not all transients analyzed are serious enough to cause a scram. The absence of multiple scram protection for several of the transients is considered acceptable because of the circumstances associated with these events. For example, the transient entitled, "Startup of a Cold Recirculation Loop," is essentially precluded by procedures and interlocks which must be violated for the incident to occur. Thus, these procedures and interlocks can be considered as backup to the single level of scram protection. Certain other transients, namely Pressure Regulator Malfunction, Inadvertent Opening of One Bypass Valve and Inadvertent Actuation of One Solenoid Relief Valve, are blowdown situations with inherent reactor power decay characteristics such that core thermal limits are not exceeded. Similarly, the Flow Controller Malfunction, resulting in zero recirculation flow, and the Recirculation Pump Trip and Stall Malfunctions are also inherent power decay situations in which the core thermal transients remain within permissible limits. The Flow Controller Malfunction resulting in maximum recirculation flow is a mild power increase transient in which the neutron flux fails to even reach the scram trip point. For the remaining transients, Failure of One Diesel Generator to Start, Power Bus Loss of Voltage and Dc Power Interruptions, there is adequate redundancy of electrical power sources and distribution systems to provide the required degree of protection. Several of the transients shown on Figure XV-1 assume 40-percent bypass capacity in their analysis. It was determined during startup testing that the actual capacity of the bypass system was slightly less than the design value. However, the shortfall in bypass capacity will not cause the transients analyzed in this chapter, which use the bypass system, to exceed any safety limits. These transients are bounded by the reload analysis, which evaluates those transient events that could result in a NMP Unit 1 UFSAR Section XV XV-3 Rev. 25, October 2017 significant reduction in MCPR. Therefore, these transients will not impact any safety limits, and will not affect the consequences of accidents which have been previously evaluated. 2.0 Methods and Assumptions Models used for evaluating the limiting transient events are described in Section IV, Reactor. A stability analysis is also described in detail in Section IV, Reactor. The pertinent reactor protective functions, and the level at which they either scram the reactor or actuate other safeguards, are given in Table XV-2. 3.0 Transient Analysis The events analyzed for each fuel cycle are summarized in the Supplemental Reload Licensing Report (SRLR)(2). 3.1 Turbine Trip Without Bypass 3.1.1 Objectives Demonstrate that the MCPR does not fall below the safety limit critical power ratio (CPR) should a turbine trip without bypass occur. 3.1.2 Assumptions and Initial Conditions 1. The reactor is at full design rating (1850 MWt) when the turbine stop valves close. 2. The turbine bypass valve system is failed in the closed position. 3. Auxiliary power is continuously supplied at rated frequency. 3.1.3 Comments The solenoid-actuated relief valve setpoints and safety valve setpoints are low enough to limit the pressure peak to below the maximum fuel external design pressure. NMP Unit 1 UFSAR Section XV XV-4 Rev. 25, October 2017 3.1.4 Results The plant response to the turbine trip without bypass transient is shown in the SRLR(2). The transient is most limiting at end-of-cycle (EOC) conditions. When the turbine stop valves close, the pressure in the pressure vessel rises. This collapses voids and leads to an increase in core power. This pushes the fuel near its MCPR limits. This transient is terminated when the reactor scram initiated by the turbine trip reduces the reactivity enough to overcome the increase in reactivity caused by the void collapse. 3.2 Loss of 100°F Feedwater Heating 3.2.1 Objective Demonstrate that MCPR does not fall below the safety limit CPR should a loss of feedwater heating occur resulting in a core power increase due to the increase in core inlet subcooling. 3.2.2 Assumptions and Initial Conditions 1. The Station is operating at full power (1850 MWt). 2. The Station is operating in the manual flow control mode. The transient is moderated by the runback in core flow if operation is in the automatic flow control mode. 3.2.3 Results Figure XV-3 shows the plant response to loss of 100°F feedwater heating transient. Feedwater heating can be lost in at least two ways: 1. Steam extraction line to heater is closed. 2. Feedwater is bypassed around heater. The first case produces a gradual cooling of the feedwater. In the second case the feedwater bypasses the heater and no heating of the feedwater is generated. In either case the reactor vessel receives cooler feedwater. The maximum number of feedwater heaters which can be tripped or bypassed by a single event represents the most severe transient for analysis considerations. This event for an instantaneous loss of the NMP Unit 1 UFSAR Section XV XV-5 Rev. 25, October 2017 feedwater heating capability of the plant causes an increase in core inlet subcooling. This increases core power due to the negative void reactivity coefficient. In automatic control some compensation of core power is realized by modulation of core flow. In any case power would increase at a very moderate rate. If power exceeded the normal full-power flow control line, the Operator would be expected to insert control rods to return the power and flow to their normal range. If this were not done the neutron flux could exceed the scram setpoint where a scram would occur. This transient is evaluated for each reload core to determine if it could potentially alter the previous cycle MCPR operating limit. If it does, the results are presented in the SRLR(2). 3.3 Feedwater Controller Failure-Maximum Demand 3.3.1 Objective Demonstrate a safe response to the malfunction which causes continuous maximum feedwater flow into the reactor. 3.3.2 Assumptions and Initial Conditions 1. Feedwater controller fails during maximum flow demand. Maximum feedwater pump discharge, based on the allowable initial pump configuration, is assumed. 2. The reactor is operating in a manual flow control mode which provides the most severe transient. 3. The transient is analyzed without credit for turbine bypass system operation. 3.3.3 Comments Feedwater could eventually be shut off by Operator action, e.g., placing the controller on manual, closing the shutoff valves or tripping the feedwater pumps. 3.3.4 Results The SRLR(2) shows the limiting transient resulting from a failure in the feedwater controller without turbine bypass. This transient causes continuous maximum feedwater flow into the NMP Unit 1 UFSAR Section XV XV-6 Rev. 25, October 2017 reactor. The influx of excess feedwater flow results in an increase in core subcooling which reduces the void fraction and, thus, induces an increase in reactor power. The excess feedwater flow also results in a rise in the reactor water level which eventually leads to turbine stop valve closure on a high reactor vessel water level signal, reactor scram, turbine trip. Reactor scram trip is actuated from main turbine stop valve position switches. Relief valves open as steam line pressures reach relief valve setpoints. 3.4 Control Rod Withdrawal Error 3.4.1 Objective Demonstrate a safe response to the control rod withdrawal error transient. 3.4.2 Assumptions and Initial Conditions 1.The reactor is operating at a power level above 75 percent of design rated power at the time the transient occurs. 2. The Reactor Operator (RO) has followed procedures and, up to the point of the withdrawal error, is in a normal mode of operation (i.e., the control rod pattern, flow setpoints, etc., are all within normal operating limits). 3. Withdrawal error occurs with the maximum worth control rod. Therefore, the maximum positive reactivity insertion will occur. 3.4.3 Comments Withdrawal of the control rod is a procedural error and not a control rod drop accident (CRDA), which is described in Section XV-C. Note: Operation in intermediate range monitor (IRM) range 10 requires total recirculation flow greater than 30 percent of rated during control rod withdrawal. This requirement is based on an analysis of the rod withdrawal error (RWE) transient which does not credit the average power range monitor (APRM) rod block system to mitigate this event. If the reload RWE transient analysis requires the APRM rod block system to minimize the CPR for this event, then the validity of the 30-percent flow NMP Unit 1 UFSAR Section XV XV-7 Rev. 25, October 2017 restriction must be reviewed and dispositioned. A Technical Specification change could be required if the analysis cannot demonstrate that 30-percent flow provides adequate margin. (See General Electric Company (GE) Report GENE-909-39-1093.) 3.4.4 Results The SRLR(2) shows the RWE transient summary and the limiting RWE rod pattern at 100 percent power and 100 percent flow (the worst-case condition). 3.5 Main Steam Line Isolation Valve Closure (With Scram) 3.5.1 Objective Demonstrate that for the design basis normal operational transient (OT) (frequent event), the relief valve capacity is sufficient to limit the pressure increase so as to prevent safety valve discharge to the containment. 3.5.2 Assumptions and Initial Conditions 1. The reactor is at the 1850 MWt power condition. 2. MSIV closure time of 4 sec and a 90-percent control blade insertion time of 3.32 sec (nominal curve). 3. The reactor scram is initiated by the valve position switches at 10 percent valve closure from full open. 4. For the purpose of this analysis the action of the emergency cooling system is ignored. 3.5.3 Comments The solenoid-actuated relief valve capacities and setpoints are determined by the turbine trip with failure of bypass transient (Section 3.1). 3.5.4 Results The pressurization transient resulting from MSIV closure with position scram is analyzed for each reload. Results of the analysis show that there is a margin greater than 60 psi between the peak calculated pressure and the lowest nominal safety valve NMP Unit 1 UFSAR Section XV XV-8 Rev. 25, October 2017 setpoint.* Therefore, the relief valve capacity is adequate to prevent lifting of the safety valves for this transient. Although the effects of the emergency cooling system are ignored in this analysis, the combined effect of the relief valves and the emergency cooling system is more than sufficient to turn the pressure transient remaining after the relief valves close for the first time. 3.6 Inadvertent Startup of Cold Recirculation Loop The following analysis provides the basis for limiting power operation to 90.5 percent of rated for partial loop operation, with the exception being if specific procedural controls as described in Technical Specification 3.1.7(e) are utilized. These procedures would preclude inadvertent cold loop startup and allow power operation at 100 percent with 1 loop isolated and 90 percent with 2 loops isolated. 3.6.1 Objective Demonstrate a safe transient for the worst procedural violation in starting up a cold recirculation loop. 3.6.2 Assumptions and Initial Conditions 1. One recirculation loop is shut down and a temperature of 150°F exists between discharge and suction valves. 2. A temperature gradient of about 380°F exists between the coolant in the lower plenum and the coolant between the discharge and suction valves of the idle recirculation loop. Note that the entire loop (from vessel suction nozzle to vessel discharge nozzle) is assumed to be at 150°F for analysis purposes. 3. Power is at a value, which at the time the analysis was performed was compatible with the recirculation flow supplied by the other four recirculation pumps, i.e., 90.5 percent of rated. 4. The four individual speed controllers of the operating loops are demanding loop flows equivalent to that obtainable at 102 percent of nominal motor generator (MG) speeds. The basis for the 60 psig margin is documented in Reference 52. NMP Unit 1 UFSAR Section XV XV-9 Rev. 25, October 2017 5. Normal scrams are in operation. 6. Procedural violations: a. As soon as the suction valve is opened, the bypass valve is opened, the pump is started and brought quickly up to the same speed demanded of the other four loops, i.e., the loop is not allowed to reverse flow and warm up. It is assumed that the recirculation pump being started does not heat up the water. b. The discharge valve is opened as soon as the pump motor has started. 3.6.3 Comment Normal procedures in starting up a cold loop are: 1. Confirm recirculation loop temperature at pump suction and discharge is at least 155°F. 2. Open the suction valve. 3. Open the bypass valve around the discharge valve and allow reverse flow to equalize temperature and observe that the loop temperature approximates that of the other loops. 4. With the loop controller on manual, start the MG set. 5. Open the discharge valve. 6. As valve opens, manually adjust speed control to match pump flow of oncoming pump with running pumps. 7. Transfer to auto, allowing the pump to operate from the master controller. 8. Normally leave the bypass valve open to reduce stress (see GE Service Information Letter SIL-104 and 104 Supplement 1). Interlocks prevent: NMP Unit 1 UFSAR Section XV XV-10 Rev. 25, October 2017 1.Starting the pump before the suction and bypass valves are opened. 2.Opening the discharge valve before the pump is started. 3.Increasing pump speed before the discharge valve is opened. 3.6.4 Results The transients resulting from the above described improper startup of a cold recirculation loop are shown on curves designated as Startup of Cold Recirculation Loop, Figure XV-8. At the start of the transient, reactor thermal power is at 90.5 percent of rated and recirculation flow is 89 percent of rated. The individual speed controllers are on manual with a demand of 102 percent of nominal generator speed. The combination of increasing recirculation flow and decreasing inlet enthalpy causes reactor power to rise such that the 122 percent of rated flux scram is reached within 6.4 sec(56). Core average surface heat flux peaks at about 99.8 percent of rated at 6.8 sec. The analysis indicates that the minimum critical heat flux ratio (MCHFR) achieved during the transient decreases, but does not fall below the value for rated power operation. The colder water in the loop does not wash completely out for about 15 sec. By this time the reactor scram has terminated the transient. 3.7 Recirculation Pump Trips 3.7.1 Objectives 1. Demonstrate that coastdown inertia is sufficient so that fuel thermal limits are not reached upon simultaneous trip of all five pumps. 2. Demonstrate a safe coastdown for other combinations of pump trips. 3.7.2 Assumptions and Initial Conditions The reactor is at design rated power (1850 MWt) with a MCHFR of about 1.9 when the pump trips occur. The use of this MCHFR herein and in succeeding transient analyses is based on the Hench-Levy correlation. NMP Unit 1 UFSAR Section XV XV-11 Rev. 25, October 2017 3.7.3 Comments 1. Loss of electric power to the drive motors of the MG sets is the definition of a pump trip. 2. For a MG set-driven pump motor, loss of field on the generator causes a faster coastdown than the loss of power to the drive motor; however, this is not as severe as the one recirculation pump stall case described under Section 3.8. Each generator is provided with its own field excitation so that simultaneous loss of more than one generator field is improbable. 3.7.4 Results The transients resulting from tripping the various combinations of recirculation pumps are shown on Figures XV-9 and XV-10. No scram trip levels are reached in any of these transients. Figure XV-10 shows the transient resulting from simultaneous trip of all five recirculation pumps. The decrease in flow causes additional void formation in the core which drops reactor power. The time constants of the fuel cause the surface heat flux to lag behind the flow decay. This mismatch between reactor thermal power and recirculation flow brings about a decrease in the critical heat flux ratio of the reactor. A MCHFR of about 1.7 was found to occur at about 1 sec. The bus arrangement permits 2 or 3 pumps to trip off with the remaining pumps still operating. These transients are less severe than the five-pump trip and result in higher MCHFRs. It is also possible to open the breakers to one MG set drive motor. The transient resulting from this accident is shown on Figure XV-9. When the single recirculation pump is tripped, the flow in that loop decreases rapidly with reverse flow occurring in about 4.75 sec. The flow in each of the four operating loops increases to a new level at which core differential, pump head and flow are again in balance. The MCHFR for this transient is much less severe than for the five-pump trip case. The two- and three-pump trip cases were not run because obviously the safety problems and MCHFR would fall between a one-pump trip and a five-pump trip transient. 3.8 Recirculation Pump Stall NMP Unit 1 UFSAR Section XV XV-12 Rev. 25, October 2017 This transient event was analyzed for the GNF2 new fuel introduction (NFI). Parameter values relate to the original analysis conditions, unless otherwise noted for the GNF2 NFI. 3.8.1 Objective Demonstrate safe coastdown for a one-pump stall (shaft seizure type). In addition, this evaluation will address reduced loop operation (RLO). SLMCPR is protected for a recirculation pump seizure in three active recirculation loops designated as the RLO limiting condition. 3.8.2 Assumptions and Initial Conditions The reactor is at the design rating power (1850 MWt) with a MCHFR of about 1.9 when one pump shaft seizes. For RLO, the event is typically evaluated at the maximum power and flow conditions allowed in RLO mode. The analysis state point with the selected power and flow conditions bounds the RLO operating domain. The event is dispositioned for BWR2 NFI since it is confirmed that the CPR of this event for a typical GNF2 application in BWR2 is negligible compared to the other more limiting events at comparable conditions. 3.8.3 Comments 1. The one-pump stall case is more severe than the case of loss of the field on one generator of the MG sets. 2. It is improbable that more than one pump can stall at the same time. 3. This event is classified in GESTAR II (Reference 14), Section S.2.2.3.4, as an accident. However, a more conservative acceptance criterion, MCPR above the SLMCPR, is used as a limit for this event, consistent with anticipated operational occurrences (AOO). Application of transient methodology and AOO criteria to this event is conservative. 3.8.4 Results NMP Unit 1 UFSAR Section XV XV-13 Rev. 25, October 2017 Figure XV-11 shows the transient resulting when the shaft of one recirculation pump seizes. No scram trip levels are reached during this transient. Reverse flow occurs in less than 1 sec. At about 5 sec this reverse flow is stabilized at about 78 percent of loop flow. The mismatch between reactor thermal power and recirculation flow for this accident is not as great as that resulting from a simultaneous trip of all five recirculation pumps. As a result of this smaller mismatch, the MCHFR for the one-pump stall transient is less severe than the five-pump trip transient. As a result of a pump seizure in the RLO operating loop, the pump speed and core flow are rapidly reduced. This is followed by a rapid reduction in neutron flux and heat flux. All key parameters settle to a new state without a scram in a relatively short time of approximately 7 sec. It is generically confirmed that the RLO recirculation pump seizure event is non-limiting for BWR2 plants to ensure that the RLO SLMCPR is protected; thus, the radiological consequences of the event are acceptable. This conclusion has been confirmed for the introduction of GNF2 fuel. 3.9 Recirculation Flow Controller Malfunction - Increase Flow 3.9.1 Objective Demonstrate a safe transient when one recirculation flow controller causes maximum increasing rate of change of pump speed. 3.9.2 Assumptions and Initial Conditions 1. The reactor is operating at partial power (reduced by flow control) when one speed controller malfunctions, causing the scoop tube positioner (for the fluid coupler in the MG set) to move at maximum speed in the direction to increase flow. 2. The scoop tube moves at a maximum rate equivalent to approximately 10 percent per second in pump speed. 3.9.3 Comments This type of malfunction should not occur simultaneously in another flow controller. A malfunction in the master flow controller would produce a less severe transient since the limit NMP Unit 1 UFSAR Section XV XV-14 Rev. 25, October 2017 on flow rate for the five speed controllers is less than the rate limit for one scoop tube positioner. 3.9.4 Results The transient resulting from this malfunction is shown on Figure XV-12. No scram trip level is reached during this transient. The starting point for this transient analysis was about 54 percent of 1850 MWt and 36 percent recirculation flow which is considered a typical low-power operating point. The malfunction in the one speed controller causes total recirculation flow to increase to about 48 percent within 9 sec. Neutron flux leads the flow and peaks at about 89 percent within 4 sec. A second peak of the same magnitude occurs at about 7.5 sec as a result of increasing feedwater flow. Peak thermal power is essentially reached at about 12 sec with a value of 73 percent of 1850 MWt. The transient terminates at this time because of a model constraint. The transient is mild, not too much more severe than the maneuver of running up in power on flow control. No thermal limits are approached during the transient. The flow in the loop in which the transient occurs increases rapidly to about 133.9 percent of rated loop flow. This flow increase from 36 percent to 133.9 percent occurs within 9 sec. 3.10 Flow Controller Malfunction - Decrease Flow 3.10.1 Objective Demonstrate a safe transient when one flow controller causes a maximum decreasing rate of change of pump speed. 3.10.2 Assumptions and Initial Conditions 1. The reactor is at design rating power (1850 MWt) when one speed controller malfunctions causing the scoop tube positioner (for the fluid coupler in the MG set) to move at maximum speed in the direction to decrease flow. 2. The scoop tube moves at a maximum rate equivalent to approximately 10 percent per second in pump speed. 3.10.3 Comments The type of malfunction assumed should not occur simultaneously in another flow controller. A malfunction in the master flow controller would produce a less severe transient since the limit NMP Unit 1 UFSAR Section XV XV-15 Rev. 25, October 2017 on flow rate for the five speed controllers is less than the rate limit for one scoop tube positioner. 3.10.4 Results The transient resulting from this malfunction is shown on Figure XV-13. No scram trip level is reached during this transient. The recirculation flow and thermal power decay and the resultant MCHFR for this transient are less severe than the tripping of one of the recirculation pumps. Reverse flow in the loop in which the controller malfunctions does not occur until about 7 sec. 3.11 Inadvertent Actuation of One Solenoid Relief Valve 3.11.1 Objectives Determine the severity of this type of blowdown and demonstrate a safe transient with regard to vessel and fuel stress limits. 3.11.2 Assumptions and Initial Conditions 1. The reactor is at design rated power (1850 MWt) operating conditions. 2. One solenoid-actuated pressure relief valve (540,910 lb/hr at 1120 psig nameplate capacity) suddenly opens and remains open. 3.11.3 Comments A permissible cooldown rate has been established in Section V-B.1.3 of the FSAR as 17.5°F/min for 10 min, followed by 100°F/hr after an appropriate soaking period. This analysis is similar for a safety valve failing to reset since the valve capacities are about the same. 3.11.4 Results The transients resulting from the inadvertent operation of one relief valve are shown on Figure XV-14. No scram trip levels are reached during this transient. The operation of the valve drops neutron power from the design rating power level of 1850 MWt to about 92 percent of this level at 4 sec. Midcore pressure drops from about 1047 psig to a low of about 1030 psig within 10 sec and then begins to rise. This NMP Unit 1 UFSAR Section XV XV-16 Rev. 25, October 2017 pressure drop is sensed by the pressure regulator and the steam flow to the turbine is decreased by the closing of the pressure regulator. Therefore, the reactor does not respond with a continuing pressure and temperature decrease. No vessel or fuel stress limits are approached. In the event the relief valve should stick open, the reactor would be shut down, depressurized and repairs made. Resetting the pressure regulator to lower settings will divert most of the steam generated to the main condenser during the depressurization. The blowdown from a stuck valve would be to the water in the suppression chamber. Total blowdown would not exceed about 150,000 lb and the suppression chamber water would undergo a temperature rise of about 25°F. Reactor coolant level would be maintained by feedwater. 3.12 Safety Valve Actuation (Overpressurization Analysis) 3.12.1 Objectives 1. To determine the adequacy of capacity and setpoint specifications. 2. To demonstrate compliance with ASME Boiler and Pressure Vessel Code, Section I-1962 and Code Case 1271N. 3. To demonstrate compliance with ASA B31.1-1955 Code for pressure piping. 3.12.2 Assumptions and Initial Conditions 1. The reactor is at design rated power (1850 MWt) operating conditions. 2. The MSIVs, all pressure-relieving valves (i.e., bypass valves, solenoid-actuated pressure relief valves, emergency cooling system), and all immediate scram actuations fail with the exception of the high neutron flux scram. 3. The setpoint of the first safety valve will prevent the pressure at the lowest point in the vessel from exceeding 1250 psig on a slow increase. 4. The safety valves are assumed to open at 103 percent of their setpoint. NMP Unit 1 UFSAR Section XV XV-17 Rev. 25, October 2017 3.12.3 Comments The principal requirements of the ASME Code are: 1. Peak vessel pressure should not exceed 110 percent of the vessel design pressure, 1250 psig. Case 1271N - Special Nuclear Ruling. 2. The trip point of at least one safety valve should be at or below the vessel design pressure. 3. The setting of the last safety valve should not exceed 103 percent of vessel design pressure for the pressure difference to the bottom of the vessel. ASA B31.1-1955 piping code permits transients up to 15 percent over primary piping design pressure (1200 psig) for 10 percent of the time. 3.12.4 Results The original analysis(1) showed that it is necessary to have a total of 16 safety valves for a design rating 1850 MWt. The supplemental analysis(2) showed that it is necessary to have a total of 9 safety valves utilizing the high flux scram for a design rating of 1850 MWt. This analysis is performed each cycle to show that the ASME Code requirements are still met. The following tabulation gives the setpoints, number, and capacity of each valve tripped at each setpoint as analyzed and specified for the design rating: Setpoint Capacity of each (psig) Number valve-lb/hr 1218 3 633,200 1227 2 637,820 1236 2 642,440 1245 1 647,070 1254 1 651,690 9 The SRLR(2) shows the transient resulting from the postulated safety valve actuation. While ordinarily the transient would be terminated by trip scram, it is assumed that all immediate scram NMP Unit 1 UFSAR Section XV XV-18 Rev. 25, October 2017 trips fail to operate (except for the high flux scram). The pressure peak obtained in the vessel is below the Code requirements of 110 percent of design pressure (1375 psig). 3.13 Feedwater Controller Malfunction (Zero Demand) 3.13.1 Objective Demonstrate that the water level drop following loss of feedwater flow does not uncover the core. 3.13.2 Assumptions and Initial Conditions 1. The reactor is at design rating power (1850 MWt). 2. The low-level scram setpoint is 1 ft below normal water level. 3. A turbine trip is initiated 5 sec after the reactor scram from low level. 3.13.3 Comments This transient is similar to the trip of both motor-driven feedwater pumps and the declutching of the main feedwater pump in relation to water level drop and other effects. 3.13.4 Results Figure XV-17 shows the first portion of the transient. As with the previous analyses of this transient in the FSAR, termination results from a transient model constraint. As before, reactor conditions at the end of the analyzed transient are used to extrapolate to the expected end conditions. The drop in level from the mismatch between steam flow from the vessel and the rapidly decreasing feedwater flow results in low-level scram at about 3.7 sec followed by the turbine trip about 5 sec later. A 5-sec delay between any scram and the turbine trip is used to isolate the two transients. At the time that the transient terminates, because of the model constraint, reactor pressure control has passed to the bypass system. The rate of decrease of level at the termination of the transient run would cause initiation of closure of the MSIVs at about 20 sec. However, the rate of pressure decrease at the turbine would produce an effective isolation prior to 20 sec NMP Unit 1 UFSAR Section XV XV-19 Rev. 25, October 2017 because the bypass valves would be closed. In fact, the closing of the bypass valves would more than likely cause isolation initiation prior to 20 sec. Conservatively assuming that the steam flow remains at the bypass capacity until 25 sec, at which time isolation is assured, less than 175 cu ft of reactor inventory is lost. Assuming this loss and accounting for no makeup from any source, condensing all steam remaining at the termination of the transient and allowing the water level to equilibrate following trip of the recirculation pumps still results in 5 ft of water above the top of the active fuel (TAF). 3.14 Turbine Trip with Partial Bypass (Low Power) 3.14.1 Objectives Demonstrate a safe abnormal operational transient within fuel thermal limits from a power level near the lower extreme of the flow control range. 3.14.2 Assumptions and Initial Conditions 1. The reactor is at a typical low power level (980 MWt) which is greater than the bypass capacity when the turbine stop valves close. 2. The turbine stop valve closure scram is initiated when the stop valves are 10 percent closed. 3. The bypass capacity is 40 percent of the turbine steam flow at the 1850 MWt level. 3.14.3 Comments The steam flow mismatch for this partial power case is only about 14 percent. The turbine stop valve closure scram is disabled only for power levels less than 45 percent. 3.14.4 Results Figure XV-18 shows the low power turbine trip transient. Because of the small mismatch between the steaming rate and the bypass capacity, the increase in pressure is quite small. NMP Unit 1 UFSAR Section XV XV-20 Rev. 25, October 2017 The action of the trip scram holds the core surface heat flux to levels below the initial value. Neutron flux peaks to only 62 percent of the 1850 MWt level. Peak pressures occur at about 2 sec as shown on Figure XV-18. The pressure regulator closes the bypass valves from about 6 sec on, such that long-term pressure control is established with the bypass valves essentially closed. 3.15 Turbine Trip with Partial Bypass (Full Power) 3.15.1 Objectives 1. Demonstrate a safe abnormal operational transient within fuel limits. 2. Verify the adequacy of the 12-sec time delay for emergency cooling system actuation at the design rating power. 3.15.2 Assumptions and Initial Conditions 1. The reactor is operating at 1850 MWt with 1030 psig vessel pressure when the turbine stop valves close. 2. The turbine stop valve closure scram is initiated before 10 percent closure. 3. The trip point for the emergency cooling system is 1080 psig while the lowest setpoint of the solenoid-actuated relief valves is 1090 psig. 4. The bypass system capacity is 40 percent of the turbine steam flow at the 1850 MWt initial power level. 3.15.3 Comments The solenoid relief valves are expected to open on this transient because of the setpoints (1090, 1095 and 1100 psig). 3.15.4 Results Figure XV-19 shows a turbine trip with partial bypass at the design rating power level (1850 MWt). The sudden closure of the turbine stop valves causes a trip scram to occur within 10 msec. NMP Unit 1 UFSAR Section XV XV-21 Rev. 25, October 2017 Neutron flux peaks at about 112 percent of the initial power level at about 0.48 sec. The core average surface heat flux does not rise above its initial value because of the action of the fast scram. Vessel dome pressure peaks at about 1116 psig while the midcore peak is about 1131 psig. These are essentially the same pressure peaks that result from a generator trip transient. Vessel pressure exceeds the 1080 psig trip point of the emergency cooling system at about 1.3 sec, indicating that the design 12-sec time delay for actuation is adequate. Full bypass flow is achieved within 0.3 sec. The valves remain open until about 18 sec when the pressure regulator demand becomes less than the combined valve capacity. Pressure exceeds the 1090 psig setpoint of the solenoid-actuated relief valves at about 1.6 sec and the valves remain open for 2.5 sec. 3.16 Inadvertent Actuation of One Bypass Valve 3.16.1 Objectives Determine the severity of this blowdown and demonstrate a safe transient with regard to vessel and fuel stress limits. 3.16.2 Assumptions and Initial Conditions 1. The reactor is at design rating power (1850 MWt). 2. One bypass valve suddenly opens at its normal opening rate (greater than that at which a bypass valve is tested) and remains open. 3.16.3 Comments A permissible cooldown rate is established in Section V-B.1.3. 3.16.4 Results Figure XV-20 shows the transient resulting from the inadvertent operation of one bypass valve. No scram trip levels are reached during this transient. The flow from one bypass valve is less than 5 percent of design steam flow and, thus, the transient is milder than the relief valve operation described in Analysis 3.11. Since the pressure at the turbine initially drops quicker than when a relief valve NMP Unit 1 UFSAR Section XV XV-22 Rev. 25, October 2017 opens, the pressure regulator responds faster. Hence, the reactor is not effected by as large a pressure disturbance. As shown on Figure XV-20, neutron flux dips to about 99 percent of the initial level at about 1 sec. Midcore pressure drops only 2 psi during the transient. Vessel and fuel thermal stresses are insignificant. 3.17 One Feedwater Pump Trip and Restart 3.17.1 Objective Demonstrate a safe transient. 3.17.2 Assumptions and Initial Conditions 1. Feedwater flow is achieved with the turbine-driven pump and two motor-driven pumps. 2. One motor-driven pump is tripped and then restarted. 3.17.3 Comments A more severe transient involving design rating power is the complete loss of feedwater as described in Analysis 3.13. 3.17.4 Results Figure XV-21 shows the transient resulting from tripping one of the feedwater pumps and restarting the pump 25 sec later. A low-level scram is not reached since the feedwater controller opens the feedwater valve too quickly to overcome the small drop in water level. The actual drop is only about 1 in. The transient is quite mild, resulting in only small parameter perturbations during the first 50 sec. 3.18 Loss of Main Condenser Vacuum Loss of condenser vacuum will initiate closure of the turbine stop valves at 20 in of Hg vacuum; for reactor power levels greater than 45 percent, this also results in an anticipatory reactor trip. Continued loss of condenser vacuum will result in turbine bypass valve closure at 10 in of Hg vacuum and main steam line (MSL) isolation at 7 in of Hg vacuum. Loss of main condenser vacuum is bounded by the analysis of turbine trip without bypass transient as described in Analysis 3.1. NMP Unit 1 UFSAR Section XV XV-23 Rev. 25, October 2017 3.19 Loss of Electrical Load (Generator Trip) 3.19.1 Objectives 1. Demonstrate a safe abnormal operational transient within fuel limits. 2. Verify the adequacy of the 12-sec time delay for emergency cooling system actuation at the design rating power level. 3.19.2 Assumptions and Initial Conditions 1. The reactor is at the 1850 MWt power condition when the generator circuit breaker trips, causing the turbine generator to overspeed. This is sensed by the turbine speed governor, which closes the control and intercept valves. Simultaneously, the initial pressure regulator sequentially opens the bypass valves. 2. The turbine stop valve closure scram is initiated before 10 percent closure. 3. The trip point for the emergency cooling system is 1080 psig, while the lowest setpoint of the solenoid-actuated relief valves is 1090 psig. 3.19.3 Comments The bypass system capacity is 40 percent of the rated steam flow at the 1850 MWt initial power level. 3.19.4 Results The transient results are shown on Figure XV-22. The sudden closure of the turbine stop valves causes a trip scram to occur within 10 msec. Neutron flux peaks at about 109 percent of the initial power level at about 0.46 sec. The core average surface heat flux does not rise above its initial value because of the action of the fast scram. Vessel dome pressure peaks at about 1109 psig while the midcore pressure peaks at about 1121 psig. These are essentially the same pressure peaks that resulted from the turbine trip with partial bypass (Figure XV-19) transient. Vessel pressure exceeds the 1080 psig trip point of the emergency cooling system at about 1.5 sec and remains above for NMP Unit 1 UFSAR Section XV XV-24 Rev. 25, October 2017 about 5.7 sec. This indicates that the design 12-sec time delay for actuation is adequate. Full bypass is achieved with 0.4 sec. The valves remain at their full open position for about 17.5 sec and then close because of the reduced demand for bypass capacity sensed by the pressure regulator. All auxiliary equipment remains energized since load transfer to the reserve buses was assumed to have occurred. 3.20 Loss of Auxiliary Power 3.20.1 Objective Demonstrate that a safe shutdown can be made with loss of auxiliary power. 3.20.2 Assumptions and Initial Conditions Loss of all auxiliary power causes loss of condenser cooling water, trip of feedwater pumps, and trip of recirculation pumps. 3.20.3 Comments A worst case for this type of transient is simulated by a simultaneous loss of condenser vacuum with resulting scram, turbine trip without bypass, and trips of the feedwater and recirculation pumps. 3.20.4 Results Figure XV-23 shows the results of the transient. Neutron flux peaks at 156 percent of the initial value at about 0.7 sec. The relief valves open in about 1.6 sec. Full flow capacity is achieved about 2.1 sec later. The relief valves continue to function until about 12.6 sec when the pressure at the location of the valves falls below the setpoint. However, heat continues to be generated after the relief valves are closed, building up pressure slowly such that the relief valves are again actuated at about 16.5 sec. The pressure drops quickly to the setpoint of the valves and again they close. The relief valves will continue to cycle as shown until the heat generation has dropped to the level where the emergency cooling system can handle it. Vessel dome pressure peaks at 1181 psig, well below the first setpoint of the safety valves. Midcore pressure peaks at about 1187 psig. NMP Unit 1 UFSAR Section XV XV-25 Rev. 25, October 2017 3.21 Pressure Regulator Malfunction 3.21.1 Objective Determine severity of this type of blowdown primarily with regard to vessel and fuel stress limits. 3.21.2 Assumptions and Initial Conditions 1. The pressure regulator fails in the direction calling for maximum turbine control and bypass valve opening. A limit equivalent to 110 percent (of the valves wide open (VWO) condition) steam flow prevents greater steam flow rates. 2. Three initial conditions were analyzed: a. 1850 MWt. b. Low power (60 percent of rated) reduced by flow control. c. Hot standby. 3.21.3 Comments A hand calculation model was used to determine the depressurization rates because of modeling limitations. 3.21.4 Results The results of these analyses are shown on Figure XV-24. No scram trip points are reached during the transients. Of the three blowdown situations analyzed, the blowdown from hot standby leads to the most rapid depressurization rate. This is to be expected since the opening of the initial pressure regulator during hot standby leads to a much greater increase in venting area through the steam line (from 0 to complete opening of the line) than in the other cases considered. The analysis indicates a 215°F decrease in coolant temperature in 5.5 min for an average rate of 39°F/min. Preliminary analyses indicate that even if the vessel temperature decreased at the same rate as the coolant, the resulting stresses would be within the limits of ASME Section III-1965, wherein a local NMP Unit 1 UFSAR Section XV XV-26 Rev. 25, October 2017 strain of about 4 percent may occur as many as ten times during lifetime. This strain limit is discussed in Section V-C.3.0. 3.22 Instrument Air Failure 3.22.1 Objective To demonstrate that instrument air failure will permit safe shutdown of the Station. 3.22.2 Assumptions and Initial Conditions Failure of the instrument air system is due to either a failure of all compressors or a large line fault. 3.22.3 Comments Section X-I.3.0 describes the redundancy of the instrument air compressors, backup of the large reservoir capacity and the service air/instrument air intertie. 3.22.4 Results The most rapid loss of instrument air would result from a rupture of the 4-in line that runs from the instrument air receiver to various locations. Rupture of this line would result in rapid loss of air to the scram pilot valves, the main steam outside isolation valves and the emergency condenser (EC) return and level control valves. Loss of air to all these would occur essentially over the same period--less than 15 sec. Loss of air to the scram pilot valves would cause a scram; loss of air to the MSIVs would cause them to close; loss of air to the EC return and level control valves would place the system in operation and begin makeup flow to the condensers. The sequence of operation of the systems specified above depends on the valve operating times. The scram valves would open 0.10 sec after loss of air. The main steam valves would close 3 to 10 sec after loss of air, depending on the closure setpoint. The EC return valves would be fully open within about 60 sec after loss of air. The initial transient would be similar to the MSIV closure in Section XV-B.3.5 of the FSAR. The pressure would reach the relief valve setpoint at about 19.6 sec. The combined action of the relief valves and reactor scram would turn the pressure transient, and the shutoff pressure of the relief valves would NMP Unit 1 UFSAR Section XV XV-27 Rev. 25, October 2017 be reached at 21.2 sec. At that time, emergency cooling system flow would be sufficient to turn the small pressure transient resulting from the closure of relief valves. When feedwater flow ceased due to the loss of both offsite power and the turbine-driven feedwater pump, a control rod drive (CRD) pump would continue to maintain level. The level control valve regulating makeup to the condensers from the EC makeup tanks would fail open upon loss of air. With both level control valves failed open, it would take 2 hr and 40 min to empty the makeup tanks. The Operator would start a condensate transfer pump to transfer water from the condensate storage tanks (CST). The valve in the discharge line from the condensate transfer system would fail closed on loss of air. To transfer condensate to the EC makeup tanks, the Operator would have to manually open the valve with a handwheel. Condensate storage, without additional makeup--at a transfer rate of 150 gpm, is adequate to make up those losses due to decay heat removal for 44 hr. The reactor temperature and pressure decay after the initial transient would be controlled by the Operator. The Operator could limit the cooldown rate to the design point of 100°F/hr by using only one EC set. This could be done by closing the EC steam supply valves for the other EC set. If the Operator desired to further slow the cooldown rate, he could cycle the supply valve to the operating EC set to achieve the desired cooldown rate. Since the shutdown cooling system flow control valves fail closed on loss of air, this system could not be used unless the valve were manually opened with the handwheel. Because the emergency cooling system with makeup to the condensers could maintain cooling for the reactor for more than 46 hr, the Operator would be able to repair the break in the air line, connect a bottled air or nitrogen supply to the shutdown system flow control valves, or manually open the valves. The Operator could then continue to cool down the reactor to shutdown conditions at the desired cooldown rate. An instrument air loss during normal Station operation or during shutdown affects the following auxiliary systems. Reactor Shutdown Cooling System There are six air-operated valves (AOVs) in the shutdown cooling system, two on each of three cooling loops. The flow control NMP Unit 1 UFSAR Section XV XV-28 Rev. 25, October 2017 valves are normally closed valves that fail closed. The other three valves are recirculation valves for pump protection that fail open. During Station operation, when the shutdown cooling system is not in operation, an air supply failure would have no effect on the shutdown cooling system. It would stay isolated. Should an air failure occur during operation of the shutdown cooling system, the system would cease to function. The three flow control valves would fail closed. Pump protection would be accomplished because the recirculation valves are designed to fail open. Handwheels on the flow control valves permit manual operation of the system. Cleanup System The three types of AOVs on the cleanup system are flow control valves, recirculation valves for pump protection, and blocking valves in drain or vent lines. The flow control valves and blocking valves fail closed. The recirculation valves fail open. Should an air failure occur when the system was shut down, nothing would happen except that the recirculation valves would open. Should an air failure occur during operation of the cleanup system, the system would cease to function, but the pumps would be protected. Control Rod Drive Hydraulic System The two AOVs in each CRD hydraulic unit are the scram inlet and outlet valves; both fail open. Should a loss of instrument air occur, low pressure in the scram valve pilot air header would cause these valves to fail open, thereby causing a reactor scram. The scram dump volume vent and drain valves are usually open, and fail closed on a loss of instrument air. However, since the reactor will have scrammed (stated above) and since these valves close on all scram signals, the loss of instrument air would not affect Station operation, except to isolate the scram dump volume in accordance with the design function. The flow control valve in operation would also close on loss of instrument air. However, since the reactor will have scrammed, this also will not affect Station operation. NMP Unit 1 UFSAR Section XV XV-29 Rev. 25, October 2017 Reactor Building Closed Loop Cooling Water (RBCLCW) System The only AOVs in the RBCLC system that affect its operation are temperature control valves. Temperature control is accomplished by two actions; control of the service water to the heat exchangers, and bypassing of the heat exchangers by the reactor building cooling water during periods of low heat load. The service water temperature control valve could fail closed. A limit stop prevents the valve from closing more than 37 percent open. The bypass flow control valve fails 39 deg open on loss of instrument air ensuring emergency essential heat removal capability. All other AOVs in this system service individual pieces of equipment, and the equipment function dictates the failure mode of these valves. Turbine Building Closed Loop Cooling Water (TBCLCW) System Operation of this system is similar to the RBCLCW system. However, in the event of loss of instrument air, the bypass flow control valve does not have a limit stop; thus, in extreme conditions total loss of turbine closed loop cooling may occur. Service Water System (SWP) The only AOVs that affect the operation of this system are the temperature control valves after the heat exchangers in the RBCLCW and TBCLCW systems. These valves could fail fully open; thus, the closed loop cooling water systems could be overcooled as a result of a loss of instrument air. Makeup Water System An air failure would cause the makeup water system (MWS) to cease functioning. An air-operated control valve in the supply to this system fails closed. Makeup water would have to come from the 40,000-gal demineralized water storage tank. Spent Fuel Storage Pool Filtering and Cooling System This system would also cease functioning if an air failure occurred. Pneumatic flow control valves and blocking valves fail closed. Fire Protection System NMP Unit 1 UFSAR Section XV XV-30 Rev. 25, October 2017 Instrument air provides supervisory air to the detector system in the five deluge valves that protect the main transformer, Station service transformer, two reserve transformers and the hydrogen storage rack. Instrument air provides air to pressurize the diesel fire pump starting air tanks. Service air provides air to keep dry pipe valves closed and supervise the preaction systems piping. Loss of instrument air will result in a loss of service air. The diesel fire pump air tanks incorporate an automatic start on low air pressure; a low air pressure alarm is provided for the dry pipe valves (which fail open), the preaction system piping and the five deluge valves. Offgas System On loss of instrument air, the blocking valve in the suction line to the two mechanical vacuum pumps will fail closed. The air valves on the discharge side of the first-stage air ejectors will also fail closed. This will prevent any release of radioactive material to the stack from the offgas (OFG) system. Control Room Ventilation System On loss of instrument air, the valves in the inlet air duct to the control room will fail closed while the valves in the emergency bypass will fail open. Startup of either of two control room emergency ventilation fans from the control room will admit outside air into the control room after passing through a high-efficiency and charcoal filter. This configuration of both emergency inlet dampers open and one emergency fan in operation has been shown to meet the requirements for control room pressure and flow (Reference 54). 3.23 Dc Power Interruptions 3.23.1 Objective Determine the capability of the plant to withstand various dc power interruptions. 3.23.2 Assumptions and Initial Conditions 1. The Station has been operating for a significant time at design power. 2. The Station is operating with all normal and reserve sources of ac power when one of the following occurs: NMP Unit 1 UFSAR Section XV XV-31 Rev. 25, October 2017 a. Interruption of power from one of two main dc battery busses. b. Interruption of power from both main dc battery busses. c. Loss of all sources of ac power coupled with complete loss of dc power. 3. A major accident such as a recirculation pipe break is not assumed to be coincident with these malfunctions. 3.23.3 Comments 1. The two dc systems within the plant are electrically independent and it is highly improbable that coincident loss of both systems could occur. 2. The assumption of complete loss of ac power would involve the loss of two 115-kV power lines, followed by the loss of the turbine generator and failure of both diesel generators to start. This series of events is considered to be highly improbable. 3.23.4 Results Failure of one main dc battery bus with normal ac power available does not represent a serious situation and Station operation continues. For loss of both main dc battery busses and all ac power available, the MSIVs close, resulting in reactor isolation and scram. The EC automatically comes into operation since the return lines to the vessel fail open upon loss of dc power. The makeup water to the EC shells is not affected. For loss of both main dc battery busses coupled with loss of all ac power, a scram results due to a number of different events. Following scram, the reactor is isolated and the EC starts up automatically due to reasons discussed above. Failure of the control to the level makeup on the condenser shells results in the level valve failing open following which manual level control is initiated. 3.24 Failure of One Diesel Generator to Start 3.24.1 Objective NMP Unit 1 UFSAR Section XV XV-32 Rev. 25, October 2017 Demonstrate adequate redundancy of the diesel generator system. 3.24.2 Assumptions and Initial Conditions A recirculation line break occurs simultaneous with loss of all ac power sources. 3.24.3 Comments 1. The diesel generator system is designed to handle the worst accident conditions. 2. Without the assumption of a major accident, the plant is safely secured by the dc battery system (see Section 3.23). 3.24.4 Results As described in Section IX, two completely independent diesel generator systems have been provided, either one of which is capable of handling the maximum load under the worst accident condition. 3.25 Power Bus Loss of Voltage 3.25.1 Objective Demonstrate that interruption of power on any 4160- or 600-V bus can be adequately handled by switching functions such that alternate sources of power are available for continued operation or an orderly shutdown can be made. 3.25.2 Assumptions and Initial Conditions Various failures of 4160- and 600-V busses throughout the plant. 3.25.3 Comments Refer to Figure IX-1 for bus designations. 3.25.4 Results Automatic transfer of the auxiliaries on power boards (PB) 11 and 12 from the normal to the reserve source PB 101 is initiated by low voltage on the auxiliary bus, generator trip, or closure of the turbine stop valves. All connected auxiliaries are NMP Unit 1 UFSAR Section XV XV-33 Rev. 25, October 2017 expected to be continued in operation by this automatic power transfer. In the event of a fault on 4160-V PB 11 or 12, the normal supply breaker to the faulted bus is tripped and transfer to reserve Station service blocked. Therefore, power is lost to all auxiliary equipment supplied by the faulted power board. Principal services interrupted are reactor recirculation pumps and feedwater. Loss of recirculation pumps is analyzed in Section 3.6 and feedwater loss in Section 3.13. The 600-V Station auxiliaries supplied by the de-energized auxiliary feeder 11 or 12 may be manually transferred to another source of power. Loss of a 4160-V auxiliary feeder de-energizes the 4160- to 600-V Station transformers. The 600-V auxiliaries involved may be manually transferred to another source of power. In any event, all important services are duplicated and complete loss of one 600-V bus does not interrupt normal operation. A fault on a 600-V power board bus section results in isolation of the faulted bus and loss of power to all connected auxiliaries. Certain 600-V auxiliaries which can be isolated from the faulted bus section can be manually transferred to another source of power. C. STANDBY SAFEGUARDS ANALYSIS 1.0 Main Steam Line Break Outside the Drywell 1.1 Identification of Causes The accident postulated is a sudden complete severance of one MSL outside the drywell with subsequent release of steam and water containing fission products to the pipe tunnel and turbine building. Venturi-type flow nozzles are utilized in each of the two MSLs inside the drywell to limit the blowdown maximum flow rate to about 200 percent of the flow rate for design rated power. Either the increased pressure drop across the limiter or local high temperature in the tunnel would initiate MSIV closure, which in turn scrams the reactor. The MSIVs are designed to close against reactor operating pressure. The initial pressure regulator senses the loss in pressure and closes the turbine inlet valves, preventing backflow from the turbine. The turbine stop valves close in approximately 2 sec, eliminating flow out of the break from the turbine steam chest side. NMP Unit 1 UFSAR Section XV XV-34 Rev. 25, October 2017 In addition to the scram from isolation valve closure, voids generated by depressurization caused by the excess flow leaving the reactor result in negative reactivity sufficient to rapidly reduce reactor power. As a final backup, low water level also scrams the reactor and low-low water level isolates the containment including the MSLs. 1.2 Accident Analysis Analysis shows that the combination of flow limiters and isolation valve closure limits reactor coolant loss to approximately 107,000 lb. The net loss in the vessel is less (101,000 lb) due to feedwater flow. Since this volume is less than the volume of water above the core, the core is not uncovered. The analysis is based on the following equipment functions and assumptions. 1. The isolation valves outside the drywell are pneumatically operated (dc solenoid-actuated) valves with adjustable closure times from 3 to 10 sec. 2. The inside isolation valves are ac motor-operated with a closure time of 10 sec. 3. The outside valves are set to close before the inside valves; assumed as 10 sec following a closure signal delay of 1 sec. 4. No credit is taken for reduction in flow as the valves close. 5. The ratio of the venturi throat area to steam line flow area is approximately 0.57. The design is such that the initial steam blowdown rate does not exceed 200 percent of the rated power flow. 6. Calculations are based on the maximum flow rate. Critical steam-water mixture flows are calculated from an energy model which compares favorably with blowdown tests.(3) Specific items considered in this analysis include: 1.2.1 Valve Closure Initiation NMP Unit 1 UFSAR Section XV XV-35 Rev. 25, October 2017 If the line break occurs close to the reactor, critical flow occurs across the venturi with a pressure differential of at least 400 psi. If the break occurs close to the turbine, the venturi pressure differential is about 185 psi. The valves are set so that initiation of closure occurs with a venturi differential pressure of 105 psi. Therefore, a break anywhere in the line initiates valve closure. The steam tunnel temperature sensors assure isolation valve closure for ruptures much smaller than that analyzed here. 1.2.2 Feedwater Flow Feedwater flow continues after the isolation valves close as reactor water level falls. The reactor core is cooled by operation of the emergency cooling and shutdown cooling systems. 1.2.3 Core Shutdown System decompression causes homogeneous boiling in saturated liquid. Increased core voids lead to rapid power reduction even before control rods are fully inserted. The energy transfer rate from the core to the moderator is determined by sensible heat stored in the fuel rods, decay heat generation, and time-dependent steam-water environment. 1.2.4 Mixture Level The decompression and core heat transfer rates determine the net vapor formation rate in the mixture. Bubbles rise by buoyancy at velocities from 0.5 to 1.0 fps relative to liquid, eventually separating from the mixture surface. However, the vapor formation rate exceeds the vapor separation rate. Consequently, mixture level rises at a velocity of 15 fps (corresponding to a 90 psi/sec decompression rate during the initial blowdown), thus rising a distance of 12 ft to spill over the dryer assembly in 0.8 sec. Thereafter, mixture blowdown continues until the isolation valves are closed. 1.2.5 Subcooled Liquid Water in the lower plenum, downcomer annulus, and recirculation loop is initially subcooled. An average uniform 19 Btu/lb subcooling is assumed for this analysis. When 880 psia saturation pressure is reached at about 9.5 sec, the subcooled water begins to flash and noticeably reduces the decompression rate. The coolant blowdown rate is not strongly affected. NMP Unit 1 UFSAR Section XV XV-36 Rev. 25, October 2017 1.2.6 System Pressure and Steam-Water Mass System pressure and associated blowdown rates are determined from mass and energy balances and saturation state relationships. The effects of core heat and feedwater are included. The blowdown characteristics change abruptly when the swelling mixture reaches the steam lines. The lower curve on Figure XV-25 shows a much faster mass loss from the system when mixture blowdown begins. The blowdown rates upon which Figure XV-25 are based are shown in Table XV-5. During the blowdown period, until high reactor water level is reached, feedwater flow is at the maximum of 2510 lb/sec. When the mixture reaches the high water level (~0.2 sec), the feedwater control valve starts to close. The valve maximum closure rate is 8 to 10 sec with feedwater flow varying linearly with time. After the valve closes, there is no feedwater makeup to the reactor. When the MSIVs close, reactor water level rapidly drops and feedwater flow is again admitted. The mass of coolant discharged to the turbine building and the net core mass loss are shown on Figure XV-25. 1.2.7 Mixture Impact Forces Mixture flow in the steam lines causes impact forces (pressures) on the isolation valves during closure. Maximum impact pressure rise for low-quality mixtures is about 200 psi, based on rigid pipe water-hammer analysis. The isolation valves are designed to close against this force. 1.2.8 Core Internal Forces System decompression and the associated expansion of steam-water causes time-dependent internal forces on various components in the vessel. Internal forces are calculated by an interconnected five-compartment model of the system. Mass, energy, flow rate, and state relationships are resolved to give continuous pressure traces for each compartment. None of the pressure differentials during blowdown impair the ability to scram or to operate the core spray system. 1.3 Radiological Effects The following analysis is based on the use of the alternative source term (AST) and RG 1.183. The operating experience at NMP Unit 1 UFSAR Section XV XV-37 Rev. 25, October 2017 Nine Mile Point Nuclear Station - Unit 1 (Unit 1) has consistently shown iodine concentrations in the range 10-2 to 10-4. These concentrations are much lower than those used in the AST accident analysis. 1.3.1 Radioactivity Releases The predominate activity in the discharged coolant would normally be N-16 which would be substantially reduced by decay before the cloud reaches the site boundary. However, this analysis is based on the assumptions described in RG 1.183 for AST, which correspond to the reactor operating with elevated coolant activity due to the presence of fuel defects. License Amendment 226 (Reference 64) revises the NMP1 licensing basis to allow for the use of release fractions listed in Tables 1 and 3 of NRC RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (ADAMS Accession No. ML003716792), for partial length fuel rods that are operating above the peak burnup limit through the end of the Operating Cycle 22. The average concentrations of iodine isotopes determined for the reactor coolant at Dresden Unit I during 1964 are given in Table XV-6. Based on the experience of similar reactors, the maximum fission product concentrations in the reactor water would occur without the cleanup system in operation when the stack offgas emission is at about 1 curie/sec after 30 min decay. The more realistic concentrations would occur with the cleanup system in operation when the stack offgas emission was at about 0.1 curie/sec after 30 min decay. For the analysis with AST, the reactor coolant activity is specified in dose equivalent (DE) I-131, which is controlled by the Technical Specifications. The AST analysis assumes the reactor coolant activity is a factor of 20 times the maximum activity concentration allowed during full power operation, which corresponds to 4.0 µCi/gm DE I-131. Table XV-6 shows the AST design basis reactor coolant activity corresponding to 4.0 µCi/gm DE I-131 used in the analysis. Assuming an 11-sec isolation valve closure time (includes circuit delays and actual closing time), a total of 2,900 curies are carried out of the break including 34 curies of I-131 and 473 curies of I-133. The noble gas activity discharged from the break is ~33 curies (no decay). NMP Unit 1 UFSAR Section XV XV-38 Rev. 25, October 2017 Coolant loss is estimated to be 107,150 lb, of which 24.5 percent is reactor steam. Measurements of halogen concentrations in the Dresden Unit I reactor water and condensate show that the steam to water halogen concentration ratio is in the range of 3 x 10-5 to 10-5. Therefore, the halogens carried out through the break are essentially those absorbed in the water. For this analysis, it is assumed that all halogens contained in the water which is vaporized on expansion to atmospheric pressure remain with the vapor. The key inputs and assumptions used in the AST dose analysis are summarized in Table XV-7a. The accident is modeled as an instantaneous release to the environment. Realistically, the activity is released to the environment over a period of time. Calculated release rates, assuming no filtration, are shown in Table XV-7b for 11-sec and 2-hr release durations. 1.3.2 Meteorology and Dose Rates Meteorology assumed for the AST MSLB accident is discussed in Section XV-C.7. The MSLB accident /Q values are shown in Table XV-7a. The accident is modeled as an instantaneous release to the environment. Activity is transported into the control room assuming an infinite exchange rate with the environment and no filtration. Doses are calculated assuming the dose conversion factors specified in Federal Guidance Reports 11 and 12. Key inputs and assumptions used in the analysis are provided in Table XV-7a. The resultant doses from the AST MSLB accident are provided in Table XV-8. The accident-specific dose acceptance criteria for the MSLB accident, assuming a pre-accident iodine activity spike and no fuel failure, are a TEDE of 25 rem at the EAB for any 2 hr, 25 rem at the outer boundary of the LPZ, and 5 rem for occupancy of the control room for the duration of the accident as specified in 10CFR50.67 and RG 1.183. The results demonstrate compliance with these acceptance criteria. 2.0 Loss-of-Coolant Accident 2.1 Introduction NMP Unit 1 UFSAR Section XV XV-39 Rev. 25, October 2017 Loss-of-coolant accident (LOCA) analyses were performed using the TRACG-LOCA methodology for all fuel types documented within Methodology TRACG-LOCA (Reference 63) is a best-estimate plus uncertainties methodology that follows Regulatory Guide 1.157. The methodology is based upon a nominal analysis coupled with a quantification of the uncertainties. TRACG-LOCA is structured using the Code Scaling, Applicability, and Uncertainty (CSAU) Evaluation Methodology approach. The methodology improves upon conservative licensing calculations by explicitly evaluating uncertainties and by eliminating excess conservatism. TRACG-LOCA improves ECCS-LOCA Reactor Vessel Model The NMP1 TRACG-LOCA vessel component is a detailed model consisting of 16 axial levels and 4 radial rings. The NMP1 vessel model includes internal components such as the reactor core, the steam separator/dryer, and control rod guide tubes. Connected components, such as the recirculation loops, feedwater lines, core spray lines, and main steam lines are also modeled. Hydraulic Calculation TRACG-LOCA uses separate field equations for the vapor and liquid phases, and it computes individual phase velocities and temperatures. Detailed calculations are included for the countercurrent flow limitation (CCFL). Break Characteristics and Flow The methodology analyzes both split and double-ended breaks. The TRACG-LOCA model accounts for break location, upstream and downstream pressures, and the break geometry. The break critical flow model was extensively validated against test data. Initial Stored Energy of the Fuel TRACG-LOCA employs realistic -state temperature distribution and stored energy. The PRIME thermal-mechanical model is utilized. Convective Heat Transfer Coefficients for BWR Rods Under Spray Cooling The TRACG-LOCA methodology applies convective heat transfer coefficients appropriate to the heat transfer regime, and these models were extensively qualified by testing. Best-estimate correlations are used for steam / droplet cooling. NMP Unit 1 UFSAR Section XV XV-40 Rev. 25, October 2017 Estimation of Overall Calculational Uncertainty TRACG-LOCA explicitly accounts for uncertainties attributable to individual models, boundary and initial conditions, and fuel behavior. Deviations between test data and the TRACG-LOCA calculations implicitly include experimental uncertainties. TRACG-LOCA calculations bound the 95th percentile value for both Peak Cladding Temperature (PCT) and cladding oxidation. Limitations and Conditions The TRACG-LOCA Safety Evaluation Report (Reference 63) imposed various limitations and conditions upon the TRACG-LOCA method. The NMP1 ECCS-LOCA evaluations were performed in accordance with all NRC-imposed limitations and conditions (Reference 62). 2.2 Input to Analysis 2.2.1 Operational and ECCS Input Parameters Table XV-9 presents the significant plant and Emergency Core Cooling System (ECCS) input parameters. 2.2.2 Single Failure Study on ECCS Manually-Controlled Electrically-Operated Valves The effects of a single failure or Operator error that causes any manually-controlled, electrically-operated valve in the ECCS to move to position that could adversely affect the ECCS has been studied. The purpose of this evaluation was to determine that any such malfunction was bounded by the LOCA calculations performed in accordance with 10 CFR 50.46. In accordance with appropriate IEEE standards, the ECCS valves are electrically assigned to different divisions of power supply. The effect of an Operator improperly actuating a single switch on the control panel is to cause only a single valve to move to an incorrect position. For the Operator error of actuating a single switch of the automatic depressurization system (ADS), the system valves are not actuated. However, the consequences of a malfunction which causes the ADS valve to inadvertently open is less severe than the recirculation pipe break. The summary of the ECCS Valve Single Failure Analysis is provided in Table XV-10. 2.2.3 Single Failure Basis NMP Unit 1 UFSAR Section XV XV-41 Rev. 25, October 2017 Reference 6 documents the NMP1 basis for core spray system single failures. The single failures considered in the LOCA analysis are shown in Table XV-11. The LOCA analysis was performed assuming one diesel generator and one EC failed, which bounds the cases shown. 2.2.4 Pipe Whip Basis Section XVI-D.2 discusses the special coping analyses performed to address the dynamic effects of high-energy line ruptures inside containment. These coping studies were not performed to demonstrate compliance with 10CFR50 Appendix A, General Design Criterion (GDC) 4, and they are not inputs to the 10CFR50.46 ECCS-LOCA analysis. However, the ability to cope with the consequences of high-energy line pipe ruptures was demonstrated using TRACG-LOCA (Reference 62) to ensure that the current fuel types/designs do not change the conclusions of those coping studies. 2.2.5 Degraded Voltage All ECCS-LOCA evaluations assume a sustained degraded voltage event coincident with the LOCA. Using bounding assumptions, an extended time delay was applied, as described in Table XV-9 for Core Spray Pump injection. The analysis results are contained in Reference 62. All cases analyzed remain below the 10 CFR 50.46(b) acceptance criteria. 2.3 This section has been deleted. 2.4 ECCS-LOCA Performance Analysis The MAPLHGR thermal limits ensure that, in the event of the worst-case LOCA, the 10 CFR 50.46(b) acceptance criteria will not be exceeded. Additionally, in compliance with the TRACG-LOCA Safety Evaluation Report (Reference 63), the maximum local oxidation is restricted to 13 % equivalent cladding reacted as calculated using Cathcart-Pawel equation. 2.4.1 Computer Code The ECCS-LOCA analysis applies the NRC-approved TRACG-LOCA methodology (Reference 63). The TRACG code is the GE-Hitachi proprietary version of the Transient Reactor Analysis Code (TRAC). NMP Unit 1 UFSAR Section XV XV-42 Rev. 25, October 2017 TRACG is a computer program used to compute the thermal hydraulic parameters and reactor response during BWR transients. The code can perform realistic threedimensional hydrodynamic calculations within the reactor vessel component, and one-dimensional calculations within other components. A full two fluid representation is employed for two-phase flow, and this allows application to transients where thermal nonequilibrium and countercurrent flow between phases is significant. TRACG is capable of modeling standard BWR fuels and advanced fuel designs-including part-length fuel rods and large water rods. 2.4.2 Description of Model Changes The general description of the TRACG-LOCA evaluation model is contained within Reference 62. There were no changes to the approved methodology during its application for the NMP1 ECCS-LOCA analysis. 2.4.3 Analysis Procedure The analysis procedure is fully described by References 62 and 63. The best-estimate NMP1 base input deck was accurately modeled using plant-specific data. A nominal break spectrum analysis was performed for the full range of possible break sizes for both the recirculation discharge and recirculation suction lines. Other break locations, including the feedwater line, the core spray line, and the main steamline were also evaluated within the break spectrum. From the break spectrum analysis, potentially limiting break locations and break sizes were identified, and uncertainty calculations were then performed for the potentially limiting small, intermediate and large break sizes. The uncertainty calculations quantify uncertainties associated with the NMP1 analysis. ECCS-LOCA phenomena and plant-specific parameters, which could have an important impact upon the results, are randomly perturbed and combined using well-established statistical techniques. For the licensing basis calculations, the uncertainty calculations provide trivariate 95/95 upper tolerance limits determined by choosing the highest-ranked order statistics for each of the three critical safety NMP Unit 1 UFSAR Section XV XV-43 Rev. 25, October 2017 parameters peak cladding oxidation, local oxidation, and core-wide metal-water reaction. The MAPLHGR limits were based upon the licensing basis uncertainty calculations, and they ensure that the 10 CFR 50.46(b) acceptance criteria are satisfied during any LOCA event. 2.4.4 Analysis Results Licensing Basis Reference 62 documents the Unit 1 ECCS-LOCA licensing basis calculations. The NMP1 Licensing Basis ECCS-LOCA analysis utilized the plant and ECCS parameters shown in Table XV-9. The break spectrum was generated for breaks ranging from 0.0028 square feet to 7.233 square feet. The break spectrum evaluations included the breaks on both discharge and suction sides of the recirculation lines, as well as the breaks in core spray, feed water and main steam lines. The break spectrum considered both split breaks and double-ended guillotine breaks. Calculations were performed at rated power and flow (100%/100%), and at the ELLLA power and flow (100%/85%), conditions. Calculations were performed for 5-loop, 4-loop, and 3-loop modes of recirculation operation. Uncertainty calculations were then performed for nine potentially limiting conditions. These uncertainty calculations included small, intermediate, and large break sizes. Rated and ELLLA power/flow conditions were included in the uncertainty calculations. Reduced recirculation loop operation was also included in the uncertainty calculations. Table XV-12 contains the ECCS-LOCA analysis results, and it demonstrates compliance with all 10 CFR 50.46(b) criteria, including the 13% restriction on maximum local oxidation. Pipe Whip Section XVI states that the recirculation line break and resultant pipe whip could damage a core spray line. This could result in the core spray system being reduced to one sparger. To account for this potential, nominal TRACG-LOCA calculations were performed that demonstrate that the 10CFR50.46(b) criteria are met. NMP Unit 1 UFSAR Section XV XV-44 Rev. 25, October 2017 These pipe whip calculations assumed: 1. Two spargers, each fed by two core spray pumps and two topping pumps, are initially available. 2. The DBA LOCA initiates from the MAPLHGR limits in the COLR. 3. Failure of a diesel generator reduces the core spray system to two spargers, each fed by one core spray pump and one topping pump. 4. The recirculation line pipe whip reduces the core spray system to one sparger fed by one core spray pump and one topping pump. 5.The remaining topping pump is turned off for diesel loading considerations leaving a single sparger fed by one core spray pump without a topping pump. The results of this nominal TRACG-LOCA analysis (Reference 62) show that the limits of 10 CFR 50.46(b), including the 13% restriction on maximum local oxidation, are met for 5-, 4- and 3-loop operation. Topping Pump Sensitivity Additional analysis (Reference 62) was performed using TRACG-LOCA to provide additional flexibility. This additional analysis assumes two core spray pump sets operate for 10 min after the LOCA occurs. After 10 min of full core spray flow, both the core spray topping pumps are shut down. The results of this analysis show that the limits of 10 CFR 50.46(b), including the 13% restriction on maximum local oxidation, are met for 5-, 4-, and 3-loop operation. This analysis allows the Operators to control flow through the core spray system and to maintain adequate NPSH to the core spray pumps by two methodsthrough the shutdown of the core spray topping pumps, in addition to the current method of throttling the core spray injection valves. Therefore, this analysis provides the Operators with more options to control core spray flow and manage emergency diesel generator loading. Relation to Technical Specifications The present core spray system Technical Specifications require two pump strings (and associated components) operable for each sparger. This requirement provides assurance that the core spray system configuration assumed in the analysis is available after the NMP Unit 1 UFSAR Section XV XV-45 Rev. 25, October 2017 worst-case single failure (loss of a diesel generator). The limiting condition of operation (LCO) allows operation for a short period of time with redundant components inoperable on the basis that a single failure need not be assumed during the short time period. The LCO does not allow the system configuration to be reduced to a single sparger level when reactor coolant temperature is above 212°F. 3.0 Refueling Accident 3.1 Identification of Causes In order to evaluate the consequences of potential accidents which could occur when the primary containment system (drywell suppression chamber) is open and only the secondary containment system is available, a number of accidents have been studied. Of the major accidents considered, an accident during the refueling with the containment head and reactor head off results in the maximum fission product release to the reactor building. The accident is assumed to occur when a fuel bundle is accidentally dropped onto the top of the core during fuel handling operations. The refueling equipment design includes a refueling platform and grapple equipped with devices especially designed for the handling and movement of the reactor fuel safely and effectively over the reactor core. Each hoist is equipped with a load limit switch and two independent travel limit switches to prevent damage to equipment due to movement in the vertical up direction. Therefore, in order for the accident to occur, the fuel assembly handle, the fuel grapple, or the grapple cable must break to allow the fuel assembly to fall at a sufficient velocity to cause fuel damage upon impact on top of the core. No nuclear excursion results from this accident even in the event the dropped fuel assembly falls into a vacant fuel location in the core. Design features include appropriate safeguards to reinforce operating procedures in preventing unsafe refueling conditions. These design features and operating procedures include the following: 1. The reactor core is designed so that it remains in a subcritical condition with one of the control rods withdrawn even if it is assumed that a fuel assembly is dropped into an empty fuel space on an otherwise fully constituted core. At least two control rods NMP Unit 1 UFSAR Section XV XV-46 Rev. 25, October 2017 adjacent to the empty fuel space must be withdrawn for a nuclear excursion to occur. 2. Procedures prohibiting rod withdrawal during movement of fuel into the reactor core are reinforced by a rod withdrawal interlock system which prevents any rod withdrawal whenever the travel limit switch on the refueling platform indicates that the platform is carrying fuel over the reactor core. This is effective when the mode switch is in the "refuel" or "shutdown" position. 3. Procedures allow multiple control rod withdrawal, but only for fuel cells with all fuel removed. These procedures strictly regulate the bypassing of single-rod-out refuel interlocks such that more than one control rod may be withdrawn. 4. Refueling procedures require source range monitor (SRM) operability and instrument observation during fuel moves. Also, the SRMs must have operable scram function capability unless all control blades are fully inserted. 5. Refueling procedures require the Station control room Operator to observe instrumentation which indicates control rod position and to be in communication with the refueling Operator during all fuel loading operations. 3.2 Accident Analysis The following analysis was generated for 8X8, 8X8R, and P8X8R fuel. However, since for the purposes of this analysis GE8X8EB fuel contains the same number and kinds of rods, and since its weight and, therefore, radiological effects are lower, this analysis bounds the effects of a dropped bundle accident involving GE8X8EB fuel. The GE11 and GNF2 fuel designs both introduced more rods per bundle as they incorporated 9x9 and 10x10 fuel rod arrays, respectively. The refueling accident results for both the 9x9 and 10x10 arrays were calculated and reported on generic bases. In both cases, the results showed more rods failed as a result of the bundle drops, but the radiological consequences were less than those reported in the following discussion. NMP Unit 1 UFSAR Section XV XV-47 Rev. 25, October 2017 Previously, the fuel handling accident was analyzed as dropping a fuel assembly onto the core from the maximum height allowed by the refueling equipment (less than 30 ft), which produces an impact velocity of 40 ft/sec. The kinetic energy acquired by the falling assembly is less than 17,000 ft-lb. This energy is dissipated in one or more impacts. The first impact is expected to dissipate most of the energy and hence produce the largest number of perforated fuel rods. In order to estimate the expected number of failed rods in each impact, an energy approach is used. The fuel assembly is expected to impact on the core at a small angle from the vertical, possibly inducing a bending mode of failure on the impacting fuel rods of the dropped assembly. The bending mode of failure is expected to absorb little energy per rod if it is assumed that each rod resists the imposed bending load by a couple consisting of two equal and opposite concentrated forces per rod. Actual perforation tests with concentrated point loads show that each rod absorbs about 1 ft-lb prior to cladding failure due to the bending mode of failure. For rods which fail due to gross compression distortion, each rod is expected to absorb about 250 ft-lb to perforation (this is based on 1 percent uniform plastic deformation of the rods). It should be noted that these energies are those expected to just perforate each rod and that, in the first impact, energy absorption much in excess of perforation energy is expected. A fuel assembly consists of about 11 percent clad structure, 17 percent other structural material and 72 percent fuel, by weight. It is conservatively assumed that none of the fuel material absorbs the kinetic energy of the fall. The energy absorption on successive impacts is estimated by consideration of a plastic impact. Conservation of momentum under a plastic impact shows that the fractional kinetic energy absorbed during impact is 1 - M1/(M1 + M2), where M1 is the impacting mass and M2 is the struck mass. Based on the fuel geometry within the core, effectively four fuel assemblies are struck by the impacting assembly. The fractional energy loss on the first impact is approximately 80 percent. The second impact is expected to be less direct; that is, the broad side of the assembly is expected to impact approximately 24 more fuel assemblies so that after the second impact only 0.008 or about 1 percent of the original kinetic energy is available for a third impact. Since 1 percent of the total kinetic energy is 170 ft-lb and a single rod is capable of NMP Unit 1 UFSAR Section XV XV-48 Rev. 25, October 2017 absorbing 250 ft-lb in compression before perforation, it is very unlikely that any rods would be perforated on a third impact. In the less likely event of the dropped assembly striking only one or two assemblies instead of four, the effective mass of the core support structure is still expected to produce a mass effect similar to the assumed four assemblies. Therefore, the energy distribution per impact is still very much higher on the first impact compared to second and possibly third impacts. The first impact dissipates 0.80 x 17,000 or 13,600 ft-lb of energy. It is assumed that 50 percent of this energy is absorbed by the dropped fuel assembly and that the remaining 50 percent is absorbed by the struck fuel assemblies in the core. Because the fuel rods of the dropped fuel assembly are susceptible to the bending mode of failure and because 1 ft-lb of energy is sufficient to cause cladding failure as a result of bending, all 63 rods of the dropped 8x8 fuel assembly and all 62 rods of the 8x8R and P8x8R assemblies are assumed to fail. Since the tie-rods of the struck fuel assemblies are more susceptible to bending failure than the other 55 or 54 fuel rods, it is assumed that they fail on the first impact. Thus, 4 x 8 = 32 tie-rods (total in four assemblies) are assumed to fail. Because the remaining fuel rods of the struck assemblies are held rigidly in place in the core, they are susceptible only to the compression mode of failure. To cause cladding failure of one fuel rod as a result of compression, 250 ft-lb of energy is required. To cause failure of all the remaining rods of the four struck assemblies, 250 x 56 x 4 or 56,000 ft-lb of energy would have to be absorbed in cladding alone. Thus, it is clear that not all the remaining fuel rods of the struck assemblies can fail on the first impact. The number of fuel rod failures caused by compression is computed as follows: 11 0.5 x 13,600 x 11 + 17 = 11 (8x8, 8x8R, P8x8R) 250 Thus, during the first impact, fuel rod failures are as follows: 8x8 8x8R/P8x8R Dropped assembly 63 rods (bending) 62 rods (bending) Struck assemblies 32 tie-rods (bending) 32 tie-rods (bending) NMP Unit 1 UFSAR Section XV XV-49 Rev. 25, October 2017 Struck assemblies 11 rods (compression) 11 rods (compression) 106 failed rods 105 failed rods Because of the less severe nature of the second impact and the distorted shape of the dropped fuel assembly, it is assumed that in only 2 of the 24 struck assemblies are the tie-rods subjected to bending failure. Thus 2 x 8 = 16 tie-rods are assumed to fail. The number of fuel rod failures caused by compression on the second impact is computed as follows: 0.19 x 17,000 x 11 2 11 + 17 = 3 (8x8, 8x8R, P8x8R) 250 Thus, during the second impact the fuel rod failures are as follows: Struck assemblies 16 tie-rods (bending) Struck assemblies 3 rods (compression) The total number of failed rods resulting from the accident is as follows: 8x8 8x8R/P8x8R GE11(8) GNF2 First impact 106 rods 105 rods 125 rods 136 equivalent full-length rods Second impact 19 rods 19 rods 15 rods 14 full-length rods Third impact 0 rods 0 rods 0 rods 0 rods 125 failed rods 124 failed rods 140 failed rods 150 equivalent full-length failed rods 3.3 Radiological Effects 3.3.1 Fission Product Releases Fission Product Release from Fuel NMP Unit 1 UFSAR Section XV XV-50 Rev. 25, October 2017 Fission product release estimates have been performed in accordance with the alternative source term (AST) methodology outlined in Regulatory Guide (RG) 1.183. License Amendment 226 (Reference 64) revises the NMP1 licensing basis to allow for the use of release fractions listed in Tables 1 and 3 of NRC RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (ADAMS Accession No. ML003716792), for partial length fuel rods that are operating above the peak burnup limit through the end of the Operating Cycle 22. The fission product source term is based on recently irradiated fuel. The source term in the fuel is increased to account for power uncertainties in accordance with the AST methodology. During the accident a number of pins are assumed to be damaged and to release the entire activity in the fuel pellet-to-clad gap region. The non-LOCA gap fractions specified in RG 1.183 are used in the analysis. Fission Product Inventory in the Reactor Building All of the noble gas fission products are assumed to be released from the reactor water to the reactor building. The halogens released are absorbed in the pool and evolve from the pool into the air to establish an overall pool decontamination effect. For the dose consequence analysis based on AST, all of the activity is conservatively assumed to be released instantaneously to the environment without filtration. Realistically, the activity will migrate from the pool water into the reactor building. As fission products are released to the reactor building, high radiation signals initiate alarms and start the emergency ventilation system. This system maintains the reactor building below atmospheric pressure and discharges a volume equivalent to 100 percent of the building volume per 24 hr through high-efficiency and charcoal filters to the stack. These safety functions are not analyzed or required for the design basis AST FHA analysis. The airborne fission product inventory in the reactor building is shown in Table XV-22. Discharge of Fission Products to Atmosphere As described above, the dose consequence analysis based on AST conservatively assumes all activity is instantaneously released directly to the environment without filtration. The corresponding design basis release rates are shown in Table NMP Unit 1 UFSAR Section XV XV-51 Rev. 25, October 2017 XV-23. The following description relates to the realistic release of fission products to the environment. Realistically, the noble gases and halogens are exhausted over a period of time from the reactor building through a dryer, a high-efficiency filter, a charcoal filter and another high-efficiency filter. Because of the relatively small heat and vapor input the building remains at relatively low temperature and humidity. The building exhaust is treated so that humidity is reduced and filter efficiency is maintained. Calculated release rates via the stack, but assuming no filtration and a 2-hr duration, are shown in Table XV-23. AST Fission Product Release Key fission product release inputs and assumptions for the AST based FHA analysis are shown in Table XV-25. The AST analysis is based upon a conservative instantaneous release model directly to the environment. The AST analysis does not credit secondary containment or the filtration provided by the reactor building emergency ventilation system (RBEVS). 3.3.2 Meteorology and Dose Rates Meteorology assumed for the AST FHA accident is discussed in Section XV-C.7. The FHA accident X/Q values are shown in Table XV-25. The accident is modeled as an instantaneous release to the environment. Activity is transported into the control room assuming an infinite exchange rate with the environment and no filtration. Doses are calculated assuming the dose conversion factors specified in Federal Guidance Reports 11 and 12. Key inputs and assumptions used in the analysis are provided in Table XV-25. The resultant doses from the AST FHA accident are provided in Table XV-24. The accident-specific dose acceptance criteria for the FHA accident are a total effective dose equivalent (TEDE) of 6.3 rem at the EAB for any 2 hr, 6.3 rem at the outer boundary of the LPZ, and 5 rem for occupancy of the control room for the duration of the accident as specified in 10CFR50.67 and RG 1.183. The results demonstrate compliance with these acceptance criteria. 4.0 Control Rod Drop Accident NMP Unit 1 UFSAR Section XV XV-52 Rev. 25, October 2017 4.1 Identification of Causes The accidental removal of a control rod from the core at a more rapid rate than that which can be achieved by the CRD system results in a power excursion. A fully-inserted control rod is assumed to become disconnected from its drive. The drive is then fully withdrawn and, subsequently, the control rod falls out of the core. The severity of the resulting excursion is reduced by strict procedural controls, supplemented by use of a rod worth minimizer (RWM). Control rod withdrawal and insertion sequences are established to assure that the maximum in-sequence individual control rod or control rod segments which are withdrawn could not be worth enough to cause the core to be more than 0.013 k supercritical if a rod drop accident were to occur. The severity is further reduced by limiting the maximum "dropout velocity" of any control rod with the rod velocity limiter. 4.2 Accident Analysis CRDA results from banked position withdrawal sequence (BPWS) plants have been statistically analyzed and documented in Reference 9. The results show that, in all cases, the peak fuel enthalpy in a rod drop accident would be much less than the 280 cal/gm design limit even with a maximum incremental rod worth corresponding to 95-percent probability at the 95-percent confidence level. Based on these results, it was proposed to the NRC, and subsequently found acceptable, to delete the CRDA from the standard GE boiling water reactor (BWR) reload package for the BPWS plants. Because of the large margin available to CRDA design limits for BPWS plants, implementation of the advanced physics methods(10) does not result in challenging the 280 cal/gm limit. Therefore, the impact of using the advanced physics methods of Reference 10, as compared to the physics methods described in Reference 11, on the generic BPWS analysis is considered negligible. 4.3 Designed Safeguards The control rod system is designed to minimize the probability of blades sticking in the core. The blades of the control rods travel in gaps between the fuel channels with approximately 1/2-in clearance and are equipped with rollers which make NMP Unit 1 UFSAR Section XV XV-53 Rev. 25, October 2017 contact with the channel walls. Since a control blade weighs approximately 220 lb, even if it separates from its drive, gravity forces would tend to make the blade follow its drive movement as if it were connected. The control rod coupling to the drive index tube significantly reduces the probability of an accidental separation of a control rod from its drive. Couplings of this design have undergone extensive tests under simulated reactor conditions and also at conditions more extreme than those expected to be encountered in reactor service. They have been operated through thousands of cycles of scram operation and a separation has never occurred. Tests have shown that the coupling will not separate when subjected to pull forces up to 20 times greater than can be applied with a CRD. Movements of the control rods, when the reactor is critical or near critical, cause changes in the neutron flux. Control rod coupling can be verified by observing the neutron flux changes during rod movement. A velocity limiter which adds substantial hydraulic drag against downward control rod movement is incorporated in the design. Testing and analysis of the velocity limiter has demonstrated a maximum rod drop velocity of 3.11 fps(53). 4.4 Procedural Safeguards Operating procedures require that control rod movements follow preplanned patterns to flatten the power distribution. A rod withdrawal procedure, incorporating the BPWS and a reduced notch worth procedure(12) (which minimizes the chance of short period scrams to an even greater extent than required for fulfilling CRDA safety requirements), forces adherence to certain constraints applied to all control rod withdrawals (and insertions) between 100-percent control rod density (all control rods inserted) and 10 percent of design rated power, in order to limit incremental control rod worths. A description of the BPWS and reduced notch worth procedure is given in References 13 and 12, respectively. Operating procedures require rod following verification checks during startup and during major rod movements, and frequent verification checks on all rods not fully inserted, to assure that any rod-from-drive separation is detected. Procedures require the full insertion of rods when following is not verified. NMP Unit 1 UFSAR Section XV XV-54 Rev. 25, October 2017 After full withdrawal from the core, a control rod sits on a seal. Procedurally, the Operator attempts to withdraw each rod to a further overtravel position. If the drive is coupled to the control rod blade, the overtravel position cannot be attained. If the drive is uncoupled, the overtravel position is reached and an indicator light warns the Operator. The drive would then be immediately reinserted to prevent possible fallout of the stuck control rod blade. This method is used on fully withdrawn control rods during reactor startup when control rod following is not verified by observing the response of the neutron flux instrumentation. 4.5 Radiological Effects The following radiological consequences are based on two release pathways, one through the mechanical vacuum pumps and the other through the main condenser based on the assumption of manual isolation of the MSIVs. Dose calculations for the CRDA do not indicate a need for mechanical vacuum pump line isolation. However, the capability was provided to automatically isolate the mechanical vacuum pump line on high radioactivity in the MSLs. As a result, dose calculations were not reevaluated based on mechanical vacuum pump line isolation; releases would be considerably less since the major pathway for radioactivity has been removed. The dose consequence analysis for the CRDA is based on the AST and RG 1.183. 4.5.1 Fission Product Releases Fission Product Release from Fuel A maximum of 10 percent of the noble gas activity and iodine activity in a fuel rod are released from the rods experiencing cladding perforation. Except for cesium and rubidium, release of solids is negligible. These conservative gap release fractions are based on AST and RG 1.183 and are the design basis for the AST dose analysis. License Amendment 226 (Reference 64) revises the NMP1 licensing basis to allow for the use of release fractions listed in Tables 1 and 3 of NRC RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (ADAMS Accession No. ML003716792), for partial length fuel rods that are operating above the peak burnup limit through the end of the Operating Cycle 22. All release fractions used in the AST dose analysis are shown in Table XV-26. Realistic estimates of maximum plenum activity based on measured activity releases from fuel with failed NMP Unit 1 UFSAR Section XV XV-55 Rev. 25, October 2017 cladding in operating reactors(14,15) are a maximum of one percent noble gas, 0.5 percent halogen and negligible solids. The fission products generated by the excursion are negligible compared to those already in the fuel due to the long-term reactor operation. Fission Product Transport Fission product transport assumptions are in accordance with AST and RG 1.183. Two cases are analyzed. The first case corresponds to a ground-level release directly from the turbine/main condenser. The second case assumes mechanical vacuum pumps are in operation resulting in an elevated release from the main stack. At hot standby, the pressure regulator maintains reactor pressure constant by bypassing steam to the condenser. A little over two full-power seconds of energy are produced in the excursion, of which less than 3 percent are released promptly and the rest released according to the relatively slow conduction heat transfer time constant of 8 to 9 sec, characteristic of UO2 fuel rods. Therefore, the increase in steam flow to the main condenser is handled by the turbine bypass system without a significant pressure transient in the reactor or in the condenser. A fraction of the fission products released from the perforated fuel rods are carried through the MSLs to the condenser. The activity is monitored in the MSLs, and alarmed in the control room upon a high activity signal. Position switches on the steam line isolation valves also actuate reactor scram when the valves are partly closed. Normally, the air ejector offgas system maintains condenser vacuum and the airborne and noncondensible fission products are carried into the offgas piping. High radiation signals isolate this piping from the stack. If the radiation is not intensive enough to cause isolation of the offgas piping and the Operator fails to isolate manually, the fission products are released to the stack, after a 30-min delay, at rates below those permitted by 10CFR20. During hot standby, the mechanical vacuum pumps are used instead of air ejectors. With the reactor isolated, the mechanical vacuum pump is not automatically isolated and continues to operate. The mechanical vacuum pump flow rate (2000 cfm) is higher than the offgas flow rate. The delay in the offgas holdup piping of about 30 min is considerably longer than the 1.75-min delay that occurs in the piping from the NMP Unit 1 UFSAR Section XV XV-56 Rev. 25, October 2017 mechanical vacuum pumps. For this reason, the accident is analyzed assuming that the high flow rate mechanical vacuum pumps are operating. Fission products released from the fuel are spread from the reactor vessel, steam line piping, turbine and condenser to the vacuum pump system. Even though some of the released noble gases are absorbed in the reactor water, all are assumed to pass to the turbine condenser system before closure of the MSIVs. The AST analysis assumes that all of the activity released from the failed fuel can become airborne. There is no assumed flashing fraction or partition fraction. All of the iodine that reaches the condenser is assumed to be either organic or elemental iodine. The release fractions are listed in Table XV-26. The partition factor (concentration in water/concentration in steam) for halogens has been measured from 3 x 104 to 105. These measurements were made at Dresden Unit I at operating power with full steam flow and voids. The measurements demonstrate that even at high flow saturated steam conditions with a high steam void content, the halogens are absorbed in the water and remain there. The halogen concentrations in the reactor coolant water and in the condenser hotwell water were measured. The ratio of halogen concentration in the reactor water to that in the condensate is the decontamination ratio, 3 x 104 to 105. The halogens are assumed to be dispersed in that amount of water which passes through the reactor during the 11 sec required for isolation valve closure. The water carry-over fraction to the turbine at rated power is normally less than 10-3. During and following the excursion, the steaming rate (including that due to decay heat and excursion energy), remains well below rated, such that the carry-over fraction is less than 10-3. However, in the AST analysis the release fractions assumed are in accordance with RG 1.183 and are listed in Table XV-26. Of the activity released to the water, 10 percent of the iodine and one percent of the other fission products are assumed to reach the condenser. Stack Release The noble gases mix with any gases in the condenser vapor space and are removed by the vacuum pump; associated stack releases are given in Tables XV-27 and XV-28. NMP Unit 1 UFSAR Section XV XV-57 Rev. 25, October 2017 The halogens reaching the condenser are absorbed in the hotwell condensate. An equilibrium is established between the halogens in this water and the halogens in the condenser vapor space. In the AST analysis, 10 percent of the iodine and one percent of the other fission products in the condenser are assumed to be available for release. The condenser vapor space is about 8.5 x 104 ft3 and the water volume is about 1.0 x 104 ft3. The condenser volume containing the source term is 5.0 x 104 ft3. The vacuum pump is operating at rated flow (2000 cfm). There are no filters in this line and holdup in the piping is only 1.75 min. 4.5.2 Meteorology and Dose Rates Activity is transported into the control room assuming an infinite exchange rate with the environment and no filtration. Doses are calculated assuming the dose conversion factors specified in Federal Guidance Reports 11 and 12. Key inputs and assumptions used in the analysis are provided in Table XV-26. The resultant doses from the AST CRDA are provided in Table XV-29. The accident-specific dose acceptance criteria for the CRDA are a TEDE of 6.3 rem at the EAB for any 2 hr, 6.3 rem at the outer boundary of the LPZ, and 5 rem for occupancy of the CR for the duration of the accident as specified in 10CFR50.67 and RG 1.183. The results demonstrate compliance with these acceptance criteria. 5.0 Containment Design Basis Accident Three containment analyses are presented in this section: 1. The original recirculation line rupture with core spray analysis is discussed in Section XV-C.5.1. This analysis evaluates the chronological events occurring and the response of Station systems, and serves as the basis for the environmental qualification of equipment located inside the drywell. 2. The original containment DBA analysis is discussed in Section XV-C.5.2. As stated in Section XV-C.5.2, its purpose is to provide the basis for the containment leakage rate limits, assuming failure of all core spray systems. NMP Unit 1 UFSAR Section XV XV-58 Rev. 25, October 2017 3. The design basis reconstitution (DBR) analysis of the long-term post-LOCA suppression chamber temperature response is discussed in Section XV-C.5.3. The DBR analysis verifies that the containment design basis heat removal requirements are satisfied at the maximum containment spray raw water temperature. The containment leakage rate design basis established in Section XV-C.5.2 is not altered by the DBR analysis. A new structural analysis was performed as a result of the Mark I Containment Program, which included the effects of loads not previously accounted for. Modifications were performed that restored the original margin of safety. The Mark I Containment Program is further discussed in Section VI-A. 5.1 Original Recirculation Line Rupture Analysis - With Core Spray 5.1.1 Purpose The full range of coolant loss accidents has been analyzed, from a small rupture where the makeup flow is greater than the coolant loss rate, to the largest, a highly improbable circumferential recirculation line break. The analysis shows that the circumferential recirculation line break (26-in diameter) results in the maximum fuel temperature and containment pressure. Because the small breaks result in longer times for blowdown and subsequent core heatup, the potential for termination of the accident by manual action is considerable. Also, the pressures in the containment systems are lower following a small break, resulting in lower leakage and dose rates. The analysis evaluates the chronological events occurring and the response of Station systems. The calculated containment conditions also serve as the basis for the environmental qualification of equipment located inside the drywell. 5.1.2 Analysis Method and Assumptions The following assumptions are made for this analysis: 1. The line is assumed to be completely severed so that coolant flows from both parts of the break. 2. The reactor is considered to be operating at 1850 MWt. NMP Unit 1 UFSAR Section XV XV-59 Rev. 25, October 2017 3. Critical two-phase flow occurs at the break(16). 4. Over the course of the reactor depressurization (14 sec), nearly all of the coolant contained in the reactor vessel blows down to the containment. 5. Although some water remains in the bottom of the reactor vessel, it is assumed all of the water inventory is discharged to the drywell as this gives a conservatively high pressure. 6. Loss of normal Station ac power. The temperature history of the core is calculated by a digital computer program. For the calculations, the core is divided into five radial zones. Each radial zone is further subdivided into five axial nodes. The fuel bundles are divided into four zones of fuel rods and are nodalized axially. With these subdivisions the temperature distribution throughout the reactor core is determined for the course of the temperature transient. Shortly after line rupture, reduced moderation due to void formation in the core halts reactor power production. A scram is initiated by either low water level in the reactor or by high pressure in the containment. When the water level reaches low-low level, containment isolation is initiated. The core spray system is initiated by low-low water level signal when reactor pressure drops below 150 psig. The combination of reactor low-low water level and high drywell pressure actuates the containment spray system. 5.1.3 Core Heat Buildup Decay heat and metal-water reaction energy are retained in the core during heatup in the form of latent and sensible heats. No radiation losses from the core to the reactor vessel or drywell are considered. Energy removal from the core is provided by the core spray system (and by hydrogen flow associated with any metal-water reaction). Radiation heat transfer from the fuel rods to the channels is considered. The computer program includes a continuous calculation of the extent and rate of metal-water reaction for all the metal surfaces within the reactor core. The metal-water reaction is defined by the expression of Baker(17). The rate of reaction is a function of both local temperature and the amount of reaction products. NMP Unit 1 UFSAR Section XV XV-60 Rev. 25, October 2017 The effective time interval over which the core is cooled during blowdown cannot be established precisely by calculation. Test data indicate, however, that a high boiling heat transfer coefficient exists during the first 2 to 6 sec of the blowdown, depending upon the power output of the particular fuel channel, the lower powered channels being cooled longest. The cooling effectiveness then falls rapidly to near zero at the end of the blowdown, 14 sec after the break. A description of the experimental program which formed the basis for the heat transfer calculations is described in APED-5458. 5.1.4 Core Spray System Subsequent to the blowdown, the core spray system cools the core. The distribution of spray flow over the top of the core and the effective cooling of the fuel bundles are predicted using the experimental correlations developed from experiments conducted specifically for this purpose. The core spray system design is based on experiments reported in APED-5458. Core spray is at full flow in less than 30 sec after severance of the line occurs. This 30-sec period includes the time required for the reactor water level to drop to the low water level scram point, as well as the time required for signal delays, emergency diesel generator startup, startup of the core spray pumps and opening of the core spray inlet valves. The results of this analysis are shown on Figure XV-56D, where the percent of fuel cladding with a temperature in excess of the plotted value is shown. During the first 30 sec after the start of the accident, the cladding temperature responds to neutron and decay heat, primary coolant cooling and the redistribution of the heat energy in the fuel rods. During the first 2 to 6 sec, water cooling keeps the cladding temperature at normal levels. The temperatures shown are the temperature of the cladding at the inside surface of the cladding. Hence, the outside surface where nucleate boiling is occurring is less than the temperature shown. As the cooling deteriorates, cladding temperatures rise rapidly due to the redistribution of the heat energy in the fuel rods. Fuel center temperatures fall because neutron power has dropped, and the temperature of the outer portion of the fuel rods rises rapidly because of the decreased cooling. The net effect is that the fuel rods tend toward a nearly uniform temperature across their diameter. With reduced cooling, the cladding temperatures approach the surface temperatures of the fuel rods. NMP Unit 1 UFSAR Section XV XV-61 Rev. 25, October 2017 Peak temperatures are effectively controlled when core spray reaches full flow. High-power density portions of the core, however, continue to heat at a reduced rate until the cladding reaches a temperature at which all decay heat can be transferred. From this time, cladding temperatures fall as decay power decreases. Until fuel cladding wetting occurs, heat is transferred from the fuel cladding by radiation to the fuel assembly flow channels, which are wetted by the core spray. Wetting of the fuel cladding by the spray begins at 1800 sec and is complete by 3600 sec. When the cladding is wetted, its temperature falls abruptly to the saturation temperature of water at containment pressure. These various processes are summarized in Table XV-29a. 5.1.5 Containment Pressure Immediately Following Blowdown The pressure in each chamber (reactor vessel, drywell and suppression chamber) depends on the mass and energy released from the reactor vessel, the rate of blowdown from the reactor, and the mass and energy blown down from the drywell to the suppression chamber. Flow losses through the vent pipes leading from the drywell to the suppression chamber, and losses through the vacuum relief valves from the suppression chamber to the drywell, are based upon a compressible flow model. The energy losses to the structures are not considered, but heat removal by the containment spray system is included in the analysis. The containment pressure history is shown on Figures XV-56E and XV-56F. The pressure response during the first 60 sec reflects the dynamic action of the pressure suppression system. During this period, the drywell pressure rises rapidly as water and steam escape from the primary system. After about 0.6 sec, the containment pressure is sufficiently high so that air, steam and liquid begin venting to the suppression chamber. The 0.6-sec delay is the result of the inertial effects of the 4-ft water leg in the vent pipes in the suppression chamber pool. After the initial 0.6-sec delay, the water and gases blow into the suppression chamber, the water being retained in the suppression chamber pool and the nitrogen from the drywell collecting above the pool. The pressure difference between the drywell and suppression chamber is the result of flow losses in the vent lines. After approximately 2 sec, the flow rate from the drywell to the suppression chamber exceeds the blowdown rate from the primary system and the drywell pressure falls. At less than 14 sec, the initial blowdown is complete, drastically reducing the steam flow to the suppression chamber such that the NMP Unit 1 UFSAR Section XV XV-62 Rev. 25, October 2017 pressure difference between the drywell and the suppression chamber is nearly zero. 5.1.6 Containment Spray One minute after the line rupture, the containment spray system is activated and there is a further drop in drywell pressure. The pressure transient following actuation of the containment spray pumps is given on Figure XV-56E with and without core spray system in operation. As the steam in the drywell is condensed by the sprays, drywell pressure drops below the pressure in the suppression chamber and the vacuum relief valves open, allowing nitrogen to purge back to the drywell. The vacuum relief valves are sized so that the pressure difference between the chambers is, at most, 0.25 psi. The decrease in pressure is resisted by the effects of decay heat, heat energy in the reactor vessel and core, and chemical energy and hydrogen from any metal-water reaction. The higher pressure transient of Figure XV-56E is due to hydrogen generation from a metal-water reaction. The heat generated from these sources is removed from the drywell airspace by the core and containment sprays. The spray water drains to the suppression chamber pool from which it is circulated through raw water cooled heat exchangers and returned to the airspace of the drywell via spray nozzles. The drywell temperature and pressure are calculated from an energy balance which considers the drywell spray, core spray, suppression chamber pool temperature and the moles of gas in the system. The containment spray water enters the vessels at the discharge temperature of the heat exchanger, and the core spray water enters the reactor at the suppression chamber pool temperature. The drywell temperature is taken to be 5°F hotter than the water discharged from the drywell to the suppression chamber. The 5°F difference represents the average discharge condition of the spray and accounts for that portion of the spray which may not be effective in cooling due to striking equipment inside the drywell. The total number of moles of noncondensible gases in the entire system (drywell and suppression chamber) is the amount of gas originally in the system plus any gas generation from metal-water reaction. It is assumed that the drywell and suppression chamber are at equal pressure due to operation of the vacuum relief system. NMP Unit 1 UFSAR Section XV XV-63 Rev. 25, October 2017 As the containment spray heat exchanger removes heat from the suppression chamber pool, system pressure falls until approximately 1000 sec due to condensation and heat removal by the containment spray system. Between 1000 sec and 3600 sec, the heat release rate from the core nearly equals the heat removed by the heat exchanger; consequently, system temperature and pressure remain nearly constant. After approximately 3600 sec, the heat exchanger removes more heat than the core is releasing and system temperature and pressure fall. The slow rate of pressure decline is caused by the slow rate of power decay. Core heat release is predicted in the core heatup calculation with the core spray system functioning. After 1 hr, all of the fuel rods are wetted and the core is quenched to saturation temperature at containment pressure. Decay heat is then the only heat source. The temperatures of the drywell and suppression chamber are shown on Figure XV-56G for this case. 5.1.7 Blowdown Effects on Core Components Pressure differences across structures and members in the reactor vessel during the blowdown are determined to assure that control rod insertion can be accomplished. Of primary concern are the forces on the control rod guide tubes below the core and on the fuel channels which guide the blades into the core. The guide tubes are designed for about 100 psi pressure differential compared to the transient peak pressure difference of about 35 psi developed during the blowdown. The transient forces last but a few seconds and are not of sufficient magnitude to interfere with rod insertion, since the large scram forces developed by the drive assure insertion should any interference develop. The most likely place at which interference might occur is in the blade space between channels. The maximum transient pressure difference across the channels varies from under 20 psi at the bottom to essentially zero at the top. This force is in a direction which could cause pinching of the blade. In that portion of the channel below the tip of a partially-inserted blade, the channel can only move until it comes into contact with the blade. This deflection is not sufficient to cause permanent distortion so that the channel springs back when the transient force decreases. Hence, no binding exists in that region of the channel except for a second or two during the transient. NMP Unit 1 UFSAR Section XV XV-64 Rev. 25, October 2017 However, for the portion of the channels above the control blades, some yielding of the channel walls occurs. The blade must then force the walls apart as it moves upward. Calculations are performed conservatively assuming that the transient peak pressure difference, which occurs across the channel at the bottom, is a steady force on the entire channel. The net normal force acting on each of the rollers is then calculated. Assuming only sliding could take place and using a coefficient of friction of unity, the total upward force required to force the walls apart is only 440 lb per blade. The CRD mechanism is characterized by high forces when scrammed. At zero reactor pressure, a drive develops a force of 6000 lb to insert the rod using the energy stored in the accumulator. The effect of the accumulator decreases as reactor pressure increases, but at a reactor pressure of 1000 psi there is still approximately 3000 lb at the beginning of the scram stroke, which is well in excess of the 400 lb calculated above. The drive can also be scrammed by reactor pressure alone. When the vessel is above 800 psig, the force exerted from this energy source is approximately 1100 lb throughout the scram. 5.1.8 Radiological Effects The radiological consequences due to a LOCA are analyzed in accordance with the AST methodology as per RG 1.183. The acceptance criteria are defined by 10CFR50.67. License Amendment 226 (Reference 64) revises the NMP1 licensing basis to allow for the use of release fractions listed in Tables 1 and 3 of NRC RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (ADAMS Accession No. ML003716792), for partial length fuel rods that are operating above the peak burnup limit through the end of the Operating Cycle 22. Although the scenario presented in this section considers core spray operation, the AST analysis performed considers core damage in excess of cladding perforation. 5.1.8.1 Fission Product Releases Release of Fission Products from the Fuel Prior to the accident the reactor is assumed to be operating at full power. The core source inventory from which the AST LOCA analysis is based is shown in Table XV-30. The release from the fuel occurs over two phases. At 2 min after initiation of the event, the gap activity is released from the fuel over a period of 30 min. The gap release is followed by the early in-vessel NMP Unit 1 UFSAR Section XV XV-65 Rev. 25, October 2017 phase which releases a significantly larger amount of activity over a period of 90 min. The fractions associated with the release from the fuel are shown in Table XV-31. The activity released from the fuel is directly released to the drywell in accordance with the AST methodology. Fission Product Release from the Drywell Directly to the Environment There are two pathways for release from the drywell directly to the environment during the LOCA using the AST methodology. The first pathway is permanent bypass leakage through several piping lines containing containment isolation valves. Leakage through these lines could bypass the reactor building and RBEVS filters resulting in a ground-level release. These lines include MSIV leakage and combined leakage from feedwater, torus vent, drywell vent, and emergency condenser vent and drain line isolation valves. The second pathway for release from the drywell directly to the environment occurs at the beginning of the event prior to establishing a sustained negative pressure in the reactor building. During this drawdown period, the release is assumed to be directly to the environment as a ground-level release. The release during drawdown is assumed to occur due to the Technical Specifications primary containment leak rate and ESF leakage. Key parameters related to bypass leakage are shown in Table XV-31. Fission Product Release from Drywell to the Reactor Building The activity released to the primary containment is subsequently assumed to be released to the reactor building at the maximum rate allowed by the Technical Specifications. Additionally, activity that is assumed to be released to the suppression pool is assumed to leak to the reactor building through ESF system leakage. Key parameters related to the leakage from the drywell to the reactor building are shown in Table XV-31. Discharge of Fission Products from Reactor Building to Atmosphere After a sustained negative pressure in the reactor building is established, activity released to the reactor building is transported to the environment by way of the RBEVS and the plant NMP Unit 1 UFSAR Section XV XV-66 Rev. 25, October 2017 stack. It is assumed that RBEVS initiates automatically and provides particulate and halogen filtration. Key parameters related to the leakage from the reactor building to the environment are shown in Table XV-31. Fission Product Transport and Removal The release fractions, leakage rates and timing are summarized in Table XV-31. The AST analysis assumes five key fission product removal mechanisms. 1. Spray removal in the drywell 2. Natural deposition in the drywell 3. Main steam line sedimentation 4. Suppression pool iodine retention 5. Filtration Prior to the onset of the gap release phase, the drywell sprays are assumed to automatically initiate. Aerosol removal due to the drywell sprays is evaluated using the proprietary STARNAUA computer code and four system-related parameters as code inputs. These four parameters are droplet size, spray flow rate, spray fall height, and sprayed volume. The maximum elemental iodine removal rate is limited to 20 hr-1 in accordance with SRP 6.5.2. The proprietary STARNAUA computer code is also used to model the natural deposition in primary containment and the gravitational settling or sedimentation credited in the main steam line bypass leakage. By crediting the liquid poison system capability to introduce sodium pentaborate into the reactor coolant within 1.5 hr to act as a buffer, the post-accident pH of the suppresion pool will remain above 7 for the duration of the accident. Therefore, AST assumptions regarding iodine retention in the suppression pool are valid and iodine re-evolution is not considered. Filtration provided by the RBEVS and CRATS is assumed to reduce the organic, elemental and particulate activity at the dose receptor locations. The filtration efficiencies assumed in the analysis are shown in Table XV-31. NMP Unit 1 UFSAR Section XV XV-67 Rev. 25, October 2017 The fission product transport is analyzed using the RADTRAD computer code and the AST methodology. The calculated activity in the drywell following the LOCA is shown in Table XV-29b. The calculated activity in the reactor building following the LOCA is shown in Table XV-29c. The activity released to the environment that results in dose to the control room personnel and offsite is shown in Table XV-29d. 5.1.8.2 Meteorology and Dose Rates Activity is transported offsite and into the control room assuming the parameters identified in Table XV-31 and using the RADTRAD computer code. Doses are calculated assuming the dose conversion factors specified in Federal Guidance Reports 11 and 12. The resultant doses from the AST LOCA are provided in Table XV-32. The accident-specific dose acceptance criteria for the LOCA are a TEDE of 25 rem at the EAB for any 2 hr, 25 rem at the outer boundary of the LPZ, and 5 rem for occupancy of the CR for the duration of the accident as specified in 10CFR50.67 and RG 1.183. The results demonstrate compliance with these acceptance criteria. 5.2 Original Containment Design Basis Accident Analysis - Without Core Spray 5.2.1 Purpose The purpose of this analysis is to provide the basis for the containment leakage rate limits, assuming a recirculation line break and failure of all core spray systems. Failure of the core spray system results in a metal-water reaction with generation of hydrogen. One of the two containment spray loops (primary or secondary) is assumed to be operative. 5.2.2 Core Heatup After the blowdown, no coolant is assumed to flow into the core except sufficient water to support the metal-water reaction. Cooling of the core is, therefore, limited to this water flow and the resulting hydrogen flow. The core cladding temperature rises (see Figure XV-57) due to decay heat to about 2000°F. At this temperature the metal-water reaction rate, as predicted by Baker(33), begins to add appreciable energy to the heatup, and as temperatures rise higher, the metal-water energy controls the heatup. The reaction terminates because the zircaloy melts, NMP Unit 1 UFSAR Section XV XV-68 Rev. 25, October 2017 runs down the hot fuel surfaces, falls through the end plate of the fuel bundle, and into the water below the core where it is quenched. Water is expected to be present at the bottom of the vessel, entering either through the core spray system, the feedwater system or CRD system. The same variables are shown on Figure XV-57 as are shown when a core spray system functions. The response of the core in both cases is the same for the first 30 sec, until the core spray system is actuated. If neither core spray system functions, clad temperatures continue to rise as shown. After about 50 sec, the temperature gradient across the fuel approaches a quasi-equilibrium and further temperature rise is due to decay heat and metal-water reaction only. As the cladding temperature continues to rise, the rate of metal-water reaction accelerates. Having conservatively assumed that sufficient water is available throughout the core to support the predicted rate of metal-water reaction, clad temperature rises at an increasing rate until the melting temperature of the cladding is reached. The molten cladding falls from the core and the heatup of that portion of the cladding is eliminated. The cumulative metal-water reaction taking place during this transient is 24.5 percent. This is the total reaction including fuel cladding and fuel channels but not including the additional reaction of the molten clad material that takes place upon quenching of the molten drops. From Figure XV-57, approximately 90 percent clad melting and 90 percent of the total metal-water reaction takes place in the first 1800 sec. The figures are used to estimate the energy and hydrogen release rates to the containment system. Calculations of the droplet diameter as the molten metal falls from the core, based on surface tension and the fuel end plate dimensions, give droplet sizes in the range of 3/8 in in diameter. Molten drop reaction rates in water temperatures of interest indicate that a reaction depth of 60 microns correlates with observed droplet reaction test data.(34) Application of the 60-micron reaction depth to the calculated mean droplet diameter results in a 4-percent reaction of the molten zirconium leaving the core. Thus, a total estimated 24.5 percent in core (Figure XV-58) and 3 percent, i.e., 0.04 x (100 - 24.5), postmelt reaction results in the total of 27.5-percent reaction in a minimum time of 30 min. Zirconium rod meltdown tests have been conducted to determine the range of droplet sizes leaving the molten core. These tests were done in a test assembly with simulated fuel end plates and NMP Unit 1 UFSAR Section XV XV-69 Rev. 25, October 2017 four induction-heated zirconium rods. One test with nine rods and an actual end plate was also conducted giving similar results. All these experiments show that a normal statistical distribution of droplet sizes ranges from a minimum of 0.137-in diameter, with a mean diameter of 0.269 in. Application of the 60-micron reaction depth to the various-sized drops shows an overall reaction of 5 percent, not appreciably different from that of 4 percent calculated initially. In addition, the tests clearly show that the molten drops are cooled by the water, thus terminating any further reaction. A description of these tests is included in APED-5454. The percent of fuel rods perforated and the percent of the fuel which is above the recrystallization temperature as a function of time are shown on Figure XV-59. 5.2.3 Containment Response One of two containment spray loops, either the primary or secondary, is assumed to function. The core heatup results are used to determine the amount of hydrogen and energy generated and released to the containment. Uniform release rates of hydrogen and energy are assumed to occur over 1800 sec. This corresponds to the time required for 90 percent of the fuel cladding to reach melting temperature (see Figure XV-57). All the hydrogen resulting from a 27.5-percent reaction of cladding and channels is assumed released. All of the resulting chemical energy as well as the decay energy and the original sensible energy in the core are also released during this time. As a result, the containment pressure rises rapidly to 25 psig at 1800 sec. After 1800 sec the hydrogen release stops and the energy release falls to decay power level. Consequently, the containment spray loop is able to quickly cool the gases in the system, sharply reducing pressure. After 2000 sec the containment pressure response is similar to the case for which the core spray system functions, except that the pressure is approximately 11 psi higher. The 11 psi difference is the result of the hydrogen generated. The temperature variations with time of the drywell and suppression chamber are shown on Figure XV-60. 5.3 Design Basis Reconstitution Suppression Chamber Heatup Analysis NMP Unit 1 UFSAR Section XV XV-70 Rev. 25, October 2017 This DBR analysis considers containment spray system operation at up to a maximum containment spray raw water temperature of 84°F. 5.3.1 Introduction The DBR program analyzed the long-term containment suppression chamber response following the containment DBA. The containment DBA, described in Section VI-B.1.2, is identified as the instantaneous rupture of the reactor coolant system (RCS) corresponding to a double-ended break of the largest pipe in the containment (coolant recirculation line). The DBR long-term containment suppression chamber response analysis(35) was performed consistent with the LOCA, described in Section XV-C.2.0, which assures that 10CFR50.46 limits are not exceeded. The Section XV-C.2.0 LOCA analysis is based on the loss of offsite power (LOOP), the single failure of one of the emergency diesel generators, and the dynamic effects of the postulated pipe break, which result in one core spray pump set available to provide core cooling. Therefore, the DBR analysis of the suppression chamber response considers core spray available and assumes less than 1 percent metal-water reaction consistent with the LOCA analysis and 10CFR50.46 limits. The design basis requirement for the containment spray system is to assure that the primary containment design pressure and temperature limits are not exceeded. In addition, the containment spray heat removal system must maintain the torus water temperature such that adequate net positive suction head (NPSH) is provided to the core spray pumps and containment spray pumps, assuming no increase in containment pressure from that present prior to the postulated LOCA. The DBR analysis of the containment heat removal design basis for the containment spray system provides a working model to assess system performance and operability since the original calculations were not available. The DBR analysis(35) methodologyproduces conservative results as compared with the original design basis analysis (Sections XV-C.2.0 and XVI-C.2.0). The DBR analysis results require that the heat removal requirements be increased, as compared with those described in Section VII-B, to assure the design basis requirements are satisfied. The increased heat removal requirements are necessary to maintain the DBR analysis conservative, as compared to the calculations described in Sections XV-C.2.0 and XVI-C.2.0. NMP Unit 1 UFSAR Section XV XV-71 Rev. 25, October 2017 The DBR analysis evaluates the containment suppression chamber response assuming the containment spray system is operated in the drywell and wetwell spray mode. Additional analyses verify that operating the containment spray system in accordance with the emergency operating procedures (EOPs) creates conditions which are bounded by the spray mode of operating the containment spray system. 5.3.2 Input to Analysis A list of significant input parameters to the DBR suppression chamber heatup analysis is presented in Table XV-32a. The method-specific inputs are discussed in Section 5.3.3.2. 5.3.3 DBR Suppression Chamber Heatup Analysis The DBR suppression chamber heatup analysis(35,60) determines the maximum torus water temperature which is expected to occur following the containment DBA. This analysis is intended to reconstitute the design basis for the containment spray system, such that the performance requirements for operation up to a maximum containment spray raw (lake) water temperature of 84°F can be assessed. This analysis does not supersede the design basis analysis discussed in Section VI-B.1.2. The Reference 35 analysis has been reanalyzed with new containment spray heat exchanger heat removal rates (K-value) in Reference 60. The revised analysis also includes different modeling assumptions on vessel pressure used for the post-blowdown break flow calculations and a different modeling of the vessel liquid and metal sensible energy. The Reference 60 analysis also includes ANS 5.1-1979 (nominal) decay heat data consistent with the Reference 35 analysis but with additional actinides and activation products included per GE Service Information Letter (SIL) 636. 5.3.3.1 Computer Codes The original calculations and/or computer analyses used to determine the design basis heat removal requirements for the containment spray system were not described in the FSAR and are not available. The DBR program chose to perform a new analysis using GE's proprietary computer code, SHEX-04. SHEX is designed to model long-term containment pressure and temperature responses to a variety of normal and abnormal operating transients, including LOCAs. SHEX-04 has been applied by NMP Unit 1 UFSAR Section XV XV-72 Rev. 25, October 2017 GE-Nuclear Energy in this type of analysis and has been reviewed and accepted by the NRC.(35) SHEX-04 evaluates the containment response by performing mass and energy balances on four main nodes: reactor pressure vessel (RPV), drywell, suppression pool and wetwell airspace. These nodes are interconnected via one or more of the auxiliarysystems; e.g., the drywell and the suppression pool are connected by the downcomers; the suppression pool and the RPV are connected by the core spray system; the drywell and wetwell airspace are connected by the wetwell to drywell vacuum breakers, etc. External mass and energy sources such as decay heat and feedwater are added to the system. The results predicted by this computer code are conservative when compared with the results of the original analysis performance assumptions based on the results of cases 1 and 2 of the Reference 35 analysis. The SHEX code has been revised for the Reference 60 analysis to allow the vessel pressure modeling described in Section 5.3.3.2. However, the methods applied for this analysis are consistent with the basic GE methodology used in long-term LOCA containment analyses. The changes to the Reference 35 analysis are for inputs and modeling assumptions and do not represent any change in the methodology. 5.3.3.2 Analysis Methods The model used in this analysis includes the RPV, drywell, wetwell (including the suppression pool), core spray system, containment spray system, feedwater, safety relief valves (SRV), main turbine, torus vents and downcomers, the drywell to wetwell vacuum breakers, and the wetwell to reactor building vacuum breakers. RPV The RPV break flow from the double-ended recirculation line break is calculated using Moody slip flow. The energy stored in the feedwater train and the energy stored in the RPV structure is added to the blowdown energy. The core spray flow is added to RPV blowdown flow to model the energy transfer. The Reference 60 re-analysis has been performed with a more realistic modeling assumption on the post-blowdown vessel pressure. The SHEX vessel fluid model used in the production NMP Unit 1 UFSAR Section XV XV-73 Rev. 25, October 2017 version of the SHEX code assumes that the vessel pressure is always equal to the saturation pressure corresponding to the vessel liquid temperature. This modeling implicitly assumes the break is always covered with water throughout the event with no inflow from the drywell. This assumption is not realistic for the DBA-LOCA. During the post-blowdown period, this modeling can result in a partial vacuum condition in the vessel which can, in turn, induce a condition whereby an unrealistic amount of the water is accumulated above the break location. This water accumulation produces the water head necessary to enable break flow out of the vessel to the drywell when the vessel is at a lower pressure than the drywell. This condition can result in an overprediction of the energy transferred from the vessel metal to the liquid. The re-analysis assumes that for a DBA-LOCA, part of the break is open to the drywell atmosphere after the initial blowdown period. The revised SHEX version simulates this assumption by ensuring that vessel pressure used in the vessel break flow calculation is no lower than the drywell pressure. The Reference 60 re-analysis has also implemented a revised approach to the vessel metal sensible energy modeling. This approach assumes that all vessel metal below 95 in above instrument zero is in contact with liquid water. This elevation corresponds to the maximum water level that is maintained by the Operator in the EOPs if the break size is insufficiently large to maintain water level below this elevation. This approach conservatively maximizes the heat transfer between the vessel metal and vessel liquid. This conservative approach also makes the transfer of metal energy to the vessel liquid effectively independent of water level. This approach ensures that the peak temperature defined for this containment accident analysis bounds the entire potential break size spectrum. Drywell/Wetwell The model of the drywell includes a holdup volume of approximately 30,200 gal. The drywell and wetwell model excludes the effect of heat transfer through the containment structures. ECCS Systems The operation of the core spray and containment spray systems is modeled consistent with the design basis requirements of these systems. Refer to Table XV-32a for a listing of the input assumptions. NMP Unit 1 UFSAR Section XV XV-74 Rev. 25, October 2017 Energy Sources In addition to the reactor coolant energy and the sensible energy of the reactor and components, the following energy sources are added: 1. Energy is added to the containment consistent from a metal-water reaction consistent with 10CFR50.46 limits (1.8 MBtus). 2. Decay heat energy consistent with an infinite exposure profile, assuming 102 percent of rated power (1887 MWt) calculated using the 1979 ANS-5.1 standard with additional actinides and activation products per GE SIL 636 is added. 5.3.3.3 Analysis Results for Containment Spray Design Basis Assumptions(60) Figure XV-60A shows the suppression chamber pool temperature heatup profile. The maximum suppression chamber pool temperature is 163.8°F, which occurs at 12,267 sec following the LOCA. The wetwell air space pressure at this time is 1.24 psi greater than the assumed initial pressure of 14.7 psia. Two containment spray pumps are assumed to auto-start when the high drywell pressure and low-low level signals occur, which is essentially at t=0. The containment spray lineup is such that the drywell and wetwell spray begin as soon as the pumps start at t=55 sec. The Operator is assumed to secure one of the two containment spray pumps at t=10 min and start one of the containment spray raw water pumps within 15 min. The containment spray heat exchanger begins to remove heat at t=15 min. The containment spray system is operated in this mode independent of drywell or wetwell pressure conditions until the temperature increase is terminated. The suppression chamber pool temperature increases from 85°F to 119°F within 30 sec. At t=190 sec, the temperature has increased to 131°F and is 142°F at t=10 min. At the 15-min mark the temperature is 146°F, at which point the containment spray heat exchangers begin to remove energy from the suppression chamber water. The temperature slowly increases and reaches the peak temperature of 163.8°F at t=12,267 sec (3 hr 24 min). NMP Unit 1 UFSAR Section XV XV-75 Rev. 25, October 2017 The drywell and wetwell pressure decreases immediately upon initiation of the sprays and is 3.5 psi within 5 min. The Operator reduces from two containment spray pumps to one at t=10 min, at which point the drywell pressure increases slightly. The pressure then decreases to about 16 psia when the raw water pump begins to cool the containment spray flow at t=15 min. The pressure then remains at about 15 psia for the duration of the event. 5.3.3.4 Analysis Results for EOP Operation Assumptions(60) Operation of the containment spray system in accordance with the EOPs requires that the Operators evaluate and perform the following actions: 1. Terminate containment spray when drywell pressure drops below 3.5 psig. 2. Initiate containment spray if the torus pressure increases above the suppression chamber spray initiation pressure. 3. Initiate torus cooling when the torus temperature is greater than 85°F. The analysis of the effect of these actions upon the peak suppression chamber temperature and pressure is performed by modeling these manual actions. Figure XV-60B shows the heatup profile. The maximum suppression chamber pool temperature is 164.9°F which occurs at t=14,288 sec. The torus airspace pressure corresponding to this peak temperature is 20.2 psia. The minimum pressure occurs at t=5 min when the drywell spray is terminated at 3.5 psig. The drywell and wetwell pressure immediately begins to increase, with the wetwell pressure reaching 20 psia at t=15 min and then slowly decreasing as heat is removed from the torus. This analysis case shows that the reduced heat removal rate associated with the torus cooling mode increases the peak temperature by about 1°F. The effect of terminating the drywell and wetwell sprays at 3.5 psig is to increase the NPSH available to the core spray and containment spray pumps, such that a less severe NPSH condition exists relative to the design basis spray mode. NMP Unit 1 UFSAR Section XV XV-76 Rev. 25, October 2017 5.3.4 Conclusions The DBR analysis results show that the peak bulk torus water temperature is between 163.8°F and 164.9°F occurring between 3 and 4 hr after the DBA event. The difference between the DBR analysis peak temperatures compared with the Section XV-C.5.2 peak of 140°F @ 1 hr is primarily because of the change in methods, not the change in maximum lake temperature assumptions. The DBR analysis could not duplicate all of the original Safety Analysis Report (SAR) methods and assumptions. Analysis of the DBR profile shows that the temperature increases to 140°F within 10 min because of the DBA blowdown. From 10 min until the peak temperature is reached, the torus heatup is governed by the heat removal capability of the containment spray system versus the heat added from decay heat. When the heat removal rate exceeds the heat added from decay heat, the temperature increase is terminated. The analysis shows that the Operator actions taken in accordance with the EOPs create conditions which assure torus pressure conditions and, in turn, improve available NPSH to the core spray pumps. The analysis also shows that the Operator actions taken in accordance with the EOPs to maintain level below 95 in ensures the peak temperature determined by the analysis bounds the entire potential break size spectrum. The DBR analysis results conclude that all the design criteria associated with maximum torus water temperature are satisfied at the calculated peak temperatures. The operability requirements imposed upon the suppression chamber (i.e., 3.5 ft minimum downcomer submergence and 85°F maximum initial torus water temperature) and upon the containment spray system (i.e., initiate containment spray raw water within 15 min) by the DBR analysis for 84°F lake temperature, limit the peak suppression chamber water temperature to less than the original heatup profile discussed in Section XV-C.5.2 when calculated on an equivalent basis.(35,60) 6.0 New Fuel Bundle Loading Error Analysis 6.1 Identification of Causes A fuel bundle loading error accident results from a misoriented or mislocated new fuel bundle in the core. This accident can only occur as a result of multiple Operator errors during NMP Unit 1 UFSAR Section XV XV-77 Rev. 25, October 2017 reloading. Either the mislocated or misoriented bundle (i.e., misoriented rotated 90 to 180) can be limiting condition for this accident. 6.2 Accident Analysis Analysis of the mislocated bundle accident is performed for reload cores where the resultant CPR response may establish the operating limit MCPR. Analysis methods for the misoriented fuel assembly are discussed in detail in Reference 38. Approval of these methods is given in Reference 39 under the stipulation that a CPR penalty of 0.02 be added for the tilted misoriented bundle. This 0.02 is added onto the calculated CPR used in determining the operating limit when utilizing this method. The fuel cladding integrity safety limit is applied to the accident results reported in the SRLR(2). The mislocated bundle analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core. A statistical comparison of actual process computer CPR data, with Haling power distribution fuel bundle CPR predictions, is performed. Using the operating data, it is possible to perform a Haling power distribution calculation for determining the fuel bundle CPRs, and then to correct them to achieve an improved prediction of the actual CPR in each fuel bundle. A detailed description of this procedure is presented in Reference 38. Fuel loading errors could result in fuel failures during Station operation. The most severe fuel failure mechanism would result from rods in a fuel assembly experiencing transition boiling resulting in clad overheating and subsequent accelerated oxidation of the cladding. The consequence of accelerated clad oxidation ultimately is cladding perforation and release of stored and generated fission products. The level of release would be in the same order of magnitude as other known fuel perforation mechanisms (PCI and hydride), and thus would not be distinguished from other fuel perforations possibly in the core. Fuel failures are detected by increased amounts of noble gas measured in the offgas system. However, Technical Specifications limit the offgas release rate (continuously monitored) and coolant activity concentrations (periodic sampling and analysis) to insure that guidelines on accidents and normal operation radiological consequences are met. Should rods in a misloaded fuel assembly fail in a more severe manner NMP Unit 1 UFSAR Section XV XV-78 Rev. 25, October 2017 than rods which fail from normal operation, the Technical Specification limits would effectively limit operation of the plant. 6.3 Safety Requirements Proper location and orientation of fuel assemblies can be readily verified by visual means and verification procedures to greatly reduce the possibility of a fuel bundle loading error. The fuel assembly loading error is classified as an accident, not a transient, so application of LHGR limits is not appropriate. The fuel bundle loading error analysis results presented in the SRLR(2) show that the MCPR will be greater than the safety limit MCPR for all exposures throughout the cycle. 7.0 Meteorological Models Used in Accident Analyses 7.1 Introduction Radiological consequences of the Unit 1 DBAs are based on atmospheric dispersion factors (/Q values). Using site meteorological data, calculations were performed to obtain the associated /Q values. These calculations used data collected by the NMPNS onsite meteorological measurements program for the 5-yr period from 1997 through 2001. 7.2 Atmospheric Dispersion Factor Calculations Meteorological data utilized for calculation of /Q values were selected from the historical record of the NMPNS meteorological monitoring program. The period from 1997 through 2001 was selected because it represents a complete and accurate data set that is representative of the site meteorological data. The data was reviewed to ensure instrumentation problems and missing or anomalous observations did not affect the validity of the data. This is consistent with the guidance in RG 1.194 that considers 5 yr of hourly observations to be representative of long-term trends. Recorded meteorological hourly average data were used to generate joint frequency distributions of wind direction, wind speed, and atmospheric stability class, in accordance with RG 1.23 and 1.145. NMP Unit 1 UFSAR Section XV XV-79 Rev. 25, October 2017 Three possible locations where accident radionuclide releases are assumed to occur are the reactor building blowout panel, the turbine building blowout panel, and the main stack. Information regarding these release points and their proximity to receptor locations is provided in Tables XV-34a and XV-34b. 7.2.1 Offsite - EAB and LPZ The computer program PAVAN is used to determine /Q values for the assessment of dose consequences of design basis accidents. The program implements the NRC guidance provided in RG 1.145. Utilizing joint frequency of occurrence distributions of wind direction, wind speed, and Pasquill atmospheric stability class, /Q values were obtained as a function of direction for various time-averaging periods at the EAB and the outer boundary of the LPZ. Analyses were made from assumed ground-level (i.e., non-elevated) releases (such as vents and building penetrations), which are less than 2.5 times the height of adjacent solid structures, and from elevated releases (i.e., stacks). Three procedures were utilized for determining /Q values for a direction-dependent approach, a direction-independent approach, and an overall site approach. The reactor building blowout panel, the turbine building blowout panel, and the main stack are the assumed accident release points. The reactor and turbine building blowout panel locations do not qualify as elevated releases as per RG 1.145. Therefore, these release points were modeled as ground-type releases. The main stack was executed as an elevated release. Source-to-receptor horizontal distances are 830 m (2,722 ft) for the EAB and 6,116 m (20,060 ft) for the LPZ. Due to the close proximity of the three release points, identical distances to the EAB and LPZ were used. NMPNS meteorological data from the 5-yr period from 1997 through 2001 was used in the analysis. Since the NMPNS meteorological data fails to provide a maximum wind speed for category 12 winds, a conservative value of 60.5 m/s was selected. The coastal sectors were not considered in determining the /Q values for the EAB and LPZ. 7.2.2 Control Room and Technical Support Center (Excluding MSLB) Control Room and TSC /Q values were calculated using ARCON96 for various source/receptor scenarios using the procedures NMP Unit 1 UFSAR Section XV XV-80 Rev. 25, October 2017 contained in RG 1.194. The scenarios were analyzed using the hourly-averaged meteorological joint wind and stability database for the 5-yr period from 1997 through 2001. All three of the assumed release points (the reactor building blowout panel, the turbine building blowout panel, and the main stack) were modeled as ground-level (vent) releases in accordance with RG 1.145 because their height is less than 2.5 times the highest adjacent structure. Geometry of the Unit 1 structures was used in the models to account for wake effects. 7.2.3 Control Room - MSLB Puff Release The MSLB accident evaluation utilizes an instantaneous "puff" release /Q. The puff release is modeled in accordance with RG 1.194, Section C.5, with the following assumed site meteorological conditions:

• Wind speed: 1 m/s toward the receptor; and
• Stability class: F The distance from the turbine building blowout panel (the assumed MSLB release point) to the control room intake is 71.9 m (236 ft). There is one air intake location. It takes approximately 136 sec for the puff to pass completely over the Unit 1 control room air intake. The control room air intake flow rate is the same during normal control room ventilation operation and emergency control room ventilation operation. As such, the control room air intake flow rate is modeled as a constant flow rate during the entire time that the MSLB puff release passes over the intake. 7.3 Summary of Results The /Q values resulting from the ARCON96 modeling analysis of each release point and meteorological database scenario for the required time intervals are shown in Tables XV-35a and XV-35b for the control room and TSC dose assessments, respectively. The /Q values for the EAB and LPZ calculated by the PAVAN modeling analysis of each release scenario are presented in Tables XV-35c and XV-35d for each of the time intervals required by RG 1.145. For the MSLB instantaneous puff release, the integrated /Q value calculated for the control room air intake is 9.979E-04 sec/m3.

NMP Unit 1 UFSAR Section XV XV-81 Rev. 25, October 2017 7.4 Exfiltration Pressure differential from the outside to the inside of the reactor building results in exfiltration from the building. If this occurs, radioactivity release bypasses the particulate and halogen removal equipment in the emergency ventilation system, and enters the atmosphere at essentially ground elevation instead of through the stack. The pressure distribution about the reactor building due to wind velocity is estimated, based on the model studies of Irminger and Nøkkentved(40) and later work by Jensen and Franck(41). Based on these pressure distributions and the design rate of leakage from the building without the ventilation system in operation, it is estimated that an exfiltration rate of 50 percent per day of the reactor building volume occurs with winds of 35 mph, increasing to 100 percent per day at 50 mph. However, due to dispersion characteristics, the worst dose at the site boundary occurs for winds of approximately 11 mph. See PHSR, Volume II, Appendix A, subsection 3.52 for building wake dilution. Based on that, the ground concentration at the site boundary in units/cc for each unit/sec emitted is 1.655 x 10-11 at 11 mph. Halogens and particulates are emitted in greater quantity than during stack release due to bypassing of the removal equipment for these fission products in the emergency ventilation system. Analyses that evaluate the effect of wind speed and the shape of the reactor building on leakage from the reactor building are presented in the following paragraphs. In any nonzero wind, some nonuniform pressure distribution is established around the building. The distribution and magnitude of the pressure depends on wind speed and direction, turbulence, and building configuration. Using the design building configuration and dimensions, an attempt was made to relate the design building to model studies conducted by others. After a suitable model was found, the worst conditions of wind turbulence and direction were determined. In determining which model most closely approximated the actual reactor building, the entire dimensions of the building above ground level were used. A height of 133 ft and wall lengths of 160 ft were used as the building dimensions. The pressure distribution about the reactor building due to wind velocity is estimated based on the model studies by Jensen and Franck.(42) The relative dimensions of the NMP Unit 1 UFSAR Section XV XV-82 Rev. 25, October 2017 model are not exactly the same, but represent the closest approach to actual building dimensions (see Figure XV-61). In calculating building exfiltration as a function of wind speed, the following basic assumptions were used: 1. Reactor building design is such that, when wind speed is 0 mph, infiltration of 100 percent of building air volume per day results when building internal pressure is a negative 0.25 in of water relative to external pressure. 2. The emergency ventilation fan maintains the above negative pressure while exhausting to the stack 100 percent of building air volume per day. 3. The referenced models represent the pressure distributions around the building. 4. Leakage paths are quite uniformly distributed. Most leakage would occur at panel joints that are distributed around the building.(43) 5. Leakage (either direction) is proportional to the differential pressure across the leakage path.(43) The south wall of the reactor building is not directly exposed to outside wind. Approximately 43 ft of the 160-ft lengths of the east and west walls are actually part of the turbine building. As shown in Figure XV-61, the portion of the structure to the left of the dotted line is not part of the reactor building. The wind pressure distributions are based on the dimensions of the entire structure. However, exfiltration or infiltration is based only on the area of the reactor building walls. Below the operating floor (el 340 ft) the metal panels are not exposed to the internal reactor building atmosphere. Below this level the building walls are concrete and the panels serve no purpose except for external building appearance. Above the operating floor the insulated panels are the sole barrier between the outside atmosphere and the building internal atmosphere. For purposes of this analysis it was assumed that leakage was negligible through the double barrier of concrete walls and metal panels below the operating floor. All leakage was assumed to be through the single-barrier metal panels above the operating floor. Similarly, the roof, with built-up five-ply felt and tar, was assumed to have negligible leakage. NMP Unit 1 UFSAR Section XV XV-83 Rev. 25, October 2017 Referenced data(44) indicate that leakage constants for metal panels of the type used are 100 to 1000 times greater than for the concrete or roof materials. The design of the emergency ventilation system is such that it is a constant-flow system if the damper settings remain unchanged. Initially, the system will be set to achieve a differential pressure of a negative 0.25 in of water relative to atmospheric static. The flow required to achieve this differential will not be more than 100 percent of building air volume per day at 0 mph. Exfiltration was analyzed for wind impinging on each wall. The most exfiltration would occur when the winds were from the north. This would result in exfiltration from the two side walls, and infiltration through the north wall and at the wall interface of the reactor and turbine buildings. Infiltration through the two walls would cause building internal pressure to be less negative and allow increased exfiltration through the two side walls. If the wind were from the south, three walls could have exfiltration and only one wall would have infiltration--the unexposed interface. Building internal pressure would have to decrease to allow enough infiltration to offset the stack release and the three-wall exfiltration. Therefore, the actual amount exfiltrated during a southerly wind would be less than for other directions. Even though more area would be available for negative wind pressure, internal pressure would decrease, resulting in less driving force for the leakage. Exfiltration for both cases have been analyzed and are presented in the following discussion. At wind speeds above 0 mph, external velocity pressure can be either negative or positive with respect to its effect on the external building surfaces. On surfaces where wind impinges, the velocity pressure is positive, so total pressure is greater than the static pressure. On surfaces where the wind does not impinge but sweeps by, the velocity pressure will be negative, so total pressure is less than static pressure. As long as the pressure on an external surface is greater than the internal surface pressure, exfiltration through the surface cannot occur. Exfiltration can occur when the wind pressure contribution becomes more negative than the internal pressure. Leakage through a surface is a function of the driving force, i.e., differential pressure, and the leakage paths. The leakage can be represented by the combination of two expressions:(45) NMP Unit 1 UFSAR Section XV XV-84 Rev. 25, October 2017 V = CC P or V = Co (P)1/2 Where: V = Volumetric leak rate (ft3/sec) CC,Co = Empirical constants for flow rates representing fluid properties and system geometries P = Differential pressure For the type of panels used, leakage is dependent on differential pressure (P) to the first power.(45) Leakage to the 0.5 power of differential pressure is insignificant. To get leakage into the units of the design basis: L = (V) (1/v) (100) (3600) (24) Where: L = Leakage (%/day) v = Building air volume (ft3) Assuming that the types and sizes of leakage paths are quite uniformly distributed over the entire exposed area of the building, varying differential pressures over various portions of the exposed area can be treated. Li = (Ai/AT) (Cc Pi) Where: Li = Leakage over some discrete area, Ai, which has a differential pressure, Pi, across it AT = Total area that is subject to leakage L = The area of each side wall, which has only a metal panel barrier, is 6850 ft2. The area of the front and back walls is 9370 ft2 each. Therefore, the total area which is subject to leakage is 32,440 ft2. NMP Unit 1 UFSAR Section XV XV-85 Rev. 25, October 2017 The leakage through the walls is balanced by out-leakage from the stack. At the design point: S = 100 = Cc (0.25) CC = 400%/day-in H20 At higher wind speeds the differential pressure is different at various points of the total leak area. As the wind sweeps air past the building, the external pressure varies as in the model studies, and internal pressure can vary to balance air that leaks in with air that is lost through the stack and walls. Pi = (PTO)i - (PTb)i Where: (PTO)i = Total pressure outside the particular area, Ai (PTb)i = Total pressure inside the building The total pressure outside the building is the algebraic sum of the static pressure plus the appropriate fraction of wind velocity pressure. Air velocity inside the building is negligible (1.85 x 10-3 ft/sec). Therefore, the total pressure inside the building is essentially static pressure. From the model pressure distributions was determined the fraction of the external area (Ai/AT) between isopressure lines. The arithmetic average of the pressure between the lines was used in determining external pressure (PTo)i. The internal pressure (PTb)i versus wind speed was determined using the leakage constant CC previously determined, the individual (PTo)i and (Ai/AT), and the stack leakage. The total exfiltration was then determined for all areas, Ai, which had a negative differential pressure (internal pressure higher than external pressure). Figure XV-62 is a curve of exfiltration versus wind speed for the worst case. The curve is based on leakage proportional to the first power of differential pressure. For comparison, leakage proportional to differential pressure to the 0.5 power was checked. For wind speeds up to 100 mph no exfiltration occurred, since internal pressure became more negative as wind speed increased. NMP Unit 1 UFSAR Section XV XV-86 Rev. 25, October 2017 Figure XV-63 is a plot of building internal pressure relative to atmospheric static pressure for the case previously analyzed and the next case, which has less exfiltration. If winds were blowing from the south, reactor building leakage would be less than Figure XV-62 indicates. For simplicity of correlation, the leakage measured during periodic testing will be assumed to be due to the most favorable direction of wind (least leakage). This more favorable case of exfiltration versus wind speed for the design building is shown on Figure XV-64. During reactor building leakage tests, the building will be required to meet the leakage of Figure XV-64. The tests will be restricted to wind speeds of less than 30 mph, since below this speed no significant exfiltration can occur. If the wind is from any direction other than the most favorable, the actual value of allowable leakage to achieve the design point would fall between the curves of Figures XV-62 and XV-64. This design point is 100 percent per day at 0.25 in of water at 0 mph. The internal pressure during the test will be required to meet the lower pressure curve of Figure XV-63, which corresponds to the leakage of Figure XV-64. If the measured pressure differential between internal static pressure and atmospheric static pressure is more negative than Figure XV-63 for a stack leakage of 100 percent per day, the building will leak less than the design point. Preferential leakage analysis of building exfiltration was also performed, using the following as possible leakage paths: 1. Panel-to-panel joints 2. Panel-to-roof joints 3. Panel-to-concrete joints Figure XV-65 shows the reactor building plan indicating these leakage paths. Details of the sections are shown on Figures XV-66 through XV-69. The only difference between this and the aforementioned analysis is the inclusion of leakage through the roof perimeter, which is at a more negative pressure distribution than any other part of the building. Building leakage is considered to be proportional to crack length. The most probable leakage would be through the panel-to-panel joints rather than through the other two paths. NMP Unit 1 UFSAR Section XV XV-87 Rev. 25, October 2017 Figures XV-70 and XV-71 show the results of this analysis and the relationship to the previous analysis, which tends to show less leakage than this revised analysis. Table XV-36 shows the length and area quantities used in the analysis. The model used for the analysis has been previously referenced in the above paragraph.(42) Another model study by Pagon(46) revealed less severe pressure distributions than the study by Jensen and Franck. For reactor building leakage tests, the assumption is that on the day of the leakage test the wind is northerly at 12 mph. The Technical Specifications require that the building internal differential pressure be at least as negative as shown on Figure XV-72, which is based on a southerly wind. This results in the most severe pressure in the reactor building, but is the case of least leakage. Since there is a northerly wind, the actual pressure curve for the design building (100 percent per day at 0 mph and 0.25 in negative pressure) exhibits less negative pressure at the same wind speeds and stack exhaust rates. Figure XV-72 shows, at the pressure curves for northerly and southerly winds, point D; that the design leakage LD = 100 percent per day. In a northerly wind the building pressure would follow curve "a" to point A where LA = 100 percent per day. However, to meet the pressure of Figure XV-72, flow must be increased to about 108 percent per day, which equals leakage at B (LB = 108 percent per day). The Technical Specifications require extrapolation back to 0 mph along curve "b" indicating LD = 108 percent per day. However, since there is a northerly wind, extrapolation back to 0 mph should be along curve "c" (parallel to curve "a"), which gives LC = 108 percent per day at 0.27 in negative pressure. But the design point is 0.25 in negative pressure. Therefore, LD = (-0.25/-0.27) LC = 100 percent per day. Thus, by extrapolating along curve "b" instead of curve "c", an error of 8 percent would be introduced if the wind were from the north. If the wind were from the south, no error would be introduced. Since wind direction probably would not remain the same throughout the test, other pressure curves between curves "a" and "b" could introduce an error ranging from 0 to 8 percent. This is conservative since the indicating leakage, as extrapolated back to the 0 mph design point, would always exceed or equal the actual leakage if the actual pressure curve for the test wind direction were used. 7.5 Secondary Containment Drawdown NMP Unit 1 UFSAR Section XV XV-88 Rev. 25, October 2017 7.5.1 Introduction The AST LOCA analysis considers the reactor building positive pressure period. This is defined as the period when a loss of offsite power (LOOP) causes a loss of reactor building negative pressure relative to the external atmospheric static pressure. The start of the emergency diesel generators followed by the start of the RBEVS returns the reactor building to a negative pressure. The time of positive pressure relative to the atmospheric status pressure is called the drawdown time. The post-LOCA primary containment leakage into the reactor building is assumed to be released directly to the environment during the drawdown period. 7.5.2 Analysis The drawdown calculations were performed using the GOTHIC 7.2a(QA) containment analysis software. In the calculations, each building's elevation was considered as well as buoyancy effects, natural circulation flow paths, and building heat sinks. The following conservative conditions were included in the analysis:

• LOOP and failure of one of the two 100-percent capacity RBEVS trains to operate (i.e., only a single RBEVS train operates).
• Maximum reactor building in-leakage allowed by Technical Specification 3.4.1 of 1,600 cfm.
• Design basis post-LOCA reactor building heat loads, including maximum post-LOCA suppression pool heatup, operation of two core spray pump sets and one containment spray pump set, heat loads from the emergency condensers on the refuel floor elevation and from the spent fuel pool (assumed to be at a constant 90°F based on manual restart of a spent fuel pool cooling pump), electrical heat loads from equipment required to operate to mitigate the LOCA, and solar heat loads.
• Winter atmospheric conditions based on onsite meteorological data collected for the 5-yr period of 1997 through 2001 (consistent with the guidance provided in RG 1.183, Appendix A, Section 4.3). The NMP Unit 1 UFSAR Section XV XV-89 Rev. 25, October 2017 use of summer conditions results in a drawdown time that is approximately one half that of the winter case and thus is less limiting. 7.5.3 Results The results of the analysis (illustrated on Figures XV-73 and XV-74) show an initial rapid rise in reactor building pressure. The reactor building pressure in the area above the refuel floor elevation (el. 340 ft) remains positive for approximately 26 min, decreases to -0.15 in WG at approximately 67 min, and reaches -0.25 in WG at approximatey 5 hr. At elevations below the refuel floor, the positive pressure times and the times to achieve -0.25 in WG are considerably shorter. For example, at the 318 ft elevation (upper), the reactor building pressure remains positive for approximately 18 min and decreases to -0.25 in WG at approximately 52 min. D. REFERENCES 1. Technical Supplement to Petition to Increase Power Level, Nine Mile Point Unit 1, April 1970. 2. "Supplemental Reload Licensing Reoprt for Nine Mile Point 1 Reload 24 Cycle 25,"002N6949, Revision 0, March 2017. 3. F. J. Moody. "Maximum Flow Rate of a Single Component, Two-Phase Mixture," Journal of Heat Transfer, Trans. ASME, Series "C" - Volume 87, February 1965, p 134. 4. Deleted. 5. Deleted. 6. P. E. Francisco. "Licensing Basis for Core Spray System Single Failure/Pipe Whip," March 28, 1990. 7. Letter, J. F. Klapproth (GE) to R. C. Jones (NRC), "Refueling Accident Analysis," dated April 24, 1992. 8. GE letter, J. F. Klapproth to USNRC, "Refueling Accident Analysis," April 24, 1992. 9. Letter, R. E. Engel (GE) to D. M. Vassallo (NRC), "Elimination of Control Rod Drop Accident Analysis for NMP Unit 1 UFSAR Section XV XV-90 Rev. 25, October 2017 Banked Position Withdrawal Sequence Plants," February 24, 1982. 10. "Steady-State Nuclear Methods," April 1985 (NEDE-30130-P-A and NEDO-30130-A). 11. J. A. Woolley. "Three Dimensional BWR Core Simulator," January 1977 (NEDO-20953A). 12. Report, "Reduced Notch Worth Procedure, Rev. 2" transmitted by letter from C. M. Richards (GE) to W. R. D'Angelo (NMPC) dated August 6, 1979. 13. C. J. Paone. "Banked Position Withdrawal Sequence," NEDO-21231, January 1977. 14. "GE Standard Application for Reactor Fuel," NEDE-24011-P-A-24, March 2017. 15. H. E. Williamson and T. C. Rowland. "Performance of Defective Fuel in the Dresden Nuclear Power Station," APED-3894, 1962. 16. F. J. Moody, op cit. 17. Louis Baker, Jr., and Louis C. Just. "Studies of Metal Water Reactions at High Temperatures III. Experimental and Theoretical Studies of the Zirconium-Water Reactions," Argonne National Laboratory, ANL 6548 (1962). 18. Deleted. 19. Deleted. 20. Deleted. 21. Deleted. 22. Deleted. 23. Deleted. 24. Deleted. 25. Deleted. 26. Deleted.

NMP Unit 1 UFSAR Section XV XV-91 Rev. 25, October 2017 27. Deleted. 28. Deleted. 29. Deleted. 30. Deleted. 31. Deleted. 32. Deleted. 33. Louis Baker, Jr., and Louis C. Just, op cit. 34. Louis Baker, Jr., and Louis C. Just, op cit, Figure 26. 35. GENE-770-91-34, "Nine Mile Point Unit 1 Pool Heatup Analysis," July 1991; NMPC Calculation SO-TORUS-M009 and NMPC Safety Evaluation 91-028. 36. Deleted. 37. Deleted. 38. Letter, R. E. Engel (GE) to D. G. Eisenhut (NRC), "Fuel Assembly Loading Error," November 30, 1977. 39. Letter, D. G. Eisenhut (NRC) to R. E. Engel (GE), MFN-200-78, May 8, 1978. 40. J. O. Irminger and Chr. Nøkkentved. "Wind Pressure on Buildings, Second Series," Copenhagen, 1936. 41. Martin Jensen and Niels Franck. "Model-Scale Tests in Turbulent Wind, Part II Phenomena Dependent on the Velocity Pressure Wind Loads on Buildings," The Danish Technical Press, Copenhagen, 1965. 42. Jensen and Franck. "Model-Scale Tests in Turbulent Wind, Part II, Phenomena Dependent on the Velocity Pressure Wind Loads on Buildings," The Danish Technical Press (Copenhagen, 1965), pp 51-54, 66. 43. R. L. Koontz, et al. "Low Pressure Containment Buildings - Components Tests and Design Data," NAA-SR-7234, Atomics International, March 15, 1963. R. L. Koontz, et al, NMP Unit 1 UFSAR Section XV XV-92 Rev. 25, October 2017 Conventional Buildings for Reactor Containment," NAA-SR-10100, Atomics International, June 15, 1965. 44. Koontz, et al. NAA-SR-10100. 45. Koontz, et al. NAA-SR-7234 and NAA-SR-10100. 46. W. W. Pagon. "Wind Tunnel Studies Reveal Pressure Distributions on Buildings," Engineering News Record, December 1934. 47. Deleted. 48. Deleted. 49. Deleted. 50. Deleted. 51. Deleted. 52. GENE-770-31-1292, Revision 2, "Engineering Report for Application of GE11 to NMP1," April 1993. 53. "Rod Drop Accident Analysis for Large Boiling Water Reactors," NEDO-10527, March 1972. 54. NMPC Modification N1-98-016. 55. GE SAFER/GESTR Analysis, GE-NE-E2100145-00-01-R1, Revision 1, "Core Spray Flow Reduction LOCA Analysis for Nine Mile Point, Unit 1," January 1999. 56. GENE-C5100196-04, "APRM Flow-Biased Trip Setpoints Long-Term Solution Option II," June 1997, NMPC Calculation No. SP-APRM-RI02A-D and NMPC Safety Evaluation No. 96-022. 57. Deleted. 58. Deleted. 59. Deleted. 60. SO-TORUS-M009, NMP-1 Torus Pool Heat-up Analysis. 61. Deleted. NMP Unit 1 UFSAR Section XV XV-93 Rev. 25, October 2017 62. "Nine Mile Point Nuclear Station Unit 1 TRACG-LOCA Loss-of-Coolant Accident Analysis for GNF2 Fuel," 002N3714, Revision 0, March 2017. 63. Licensing Topical Report, "TRACG Application for Emergency Core Cooling Systems/Loss-of-Coolant Accident Analysis for BWR/2-6," NEDE-33005P-A, Revision 1, February 2017. 64. Nine Mile Point Nuclear Station, Unit 1 Issuance of Amendment RE: Partial Length Rod Burnup (CAC No. MF9046), March 9, 2017, ADAMS Accession Number: ML17055A451. NMP Unit 1 UFSAR Section XV XV-94 Rev. 25, October 2017 TABLE XV-1 THIS TABLE HAS BEEN DELETED NMP Unit 1 UFSAR Section XV XV-95 Rev. 25, October 2017 TABLE XV-2 TRIP POINTS FOR PROTECTIVE FUNCTIONS Reactor High Pressure 1080 psig Reactor High Flux 122 percent Reactor Low Water Level 53 in (indicator scale) Position-Main Steam Line 10 percent valve closure Isolation Valve from full open Actuation of Emergency 1080 psig with 12-sec time Condensers delay or reactor low-low water level (5 in indicator scale) with 12-sec time delay Open Solenoid Actuated 1090, 1095 and 1100 psig Relief Valves Safety Valves See Section XV-B.3.12 NMP Unit 1 UFSAR Section XV XV-96 Rev. 25, October 2017 TABLE XV-3 THIS TABLE HAS BEEN DELETED NMP Unit 1 UFSAR Section XV XV-97 Rev. 25, October 2017 TABLE XV-4 THIS TABLE HAS BEEN DELETED. NMP Unit 1 UFSAR Section XV XV-98 Rev. 25, October 2017 TABLE XV-5 BLOWDOWN RATES Average Average Blowdown Rate Enthalpy Time Flash (lb/sec) (Btu/lb) 0 - 1.8 sec --- 3,360 1190 1.8 - 9.5 --- 11,000 675 9.5 - 11.0 Subcooled 11,000 670 Water NMP Unit 1 UFSAR Section XV XV-99 Rev. 25, October 2017 TABLE XV-6 REACTOR COOLANT CONCENTRATIONS (Ci/gm) NMP Unit 1 Dresden I Cleanup System AST Isotopes 1964 in Operation MSLB Analysis I-131 .025 .03 0.7 I-133 .100 .20 9.7 I-132 { } { } 13.0 { } { } I-134 { .25 } { 1.33 } 23.6 { } { } I-135 { } { } 9.7 Other Fission .25 .14 .25 (Alkali Products Metals) 13.5 (Noble Gases) NMP Unit 1 UFSAR Section XV XV-100 Rev. 25, October 2017 TABLE XV-7 THIS TABLE HAS BEEN DELETED NMP Unit 1 UFSAR Section XV XV-101 Rev. 25, October 2017 TABLE XV-7a MSLB ACCIDENT ANALYSIS INPUTS AND ASSUMPTIONS Parameter Reactor Coolant Activity, DE I-131 Equilibrium Pre-accident Spike Failed Fuel Break Isolation Time Mass Released Steam Fraction Holdup Credit Radioactive Decay Fission Product Removal Dose Conversion Factors Control Room Intake Flow Control Room Volume Filtration Atmospheric Dispersion, /Q EAB LPZ CR Value 0.2 µCi/gm 4.0 µCi/gm None 11 sec 107,150 lbm 24.5% None None None FGR 11 & 12 Infinite Exchange Rate 135,000 ft3 None 1.90 x 10-4 sec/m3 1.63 x 10-5 sec/m3 9.98 x 10-4 sec/m3 NMP Unit 1 UFSAR Section XV XV-102 Rev. 25, October 2017 TABLE XV-7b MSLB ACCIDENT RELEASE RATES Release Duration 11 sec 2 hr Iodines (Ci/sec) 250 0.383 Noble Gases (Ci/sec) 2.99 0.00457 Alkali Metals (Ci/sec) 11.1 0.0170 NMP Unit 1 UFSAR Section XV XV-103 Rev. 25, October 2017 TABLE XV-8 MAIN STEAM LINE BREAK ACCIDENT DOSES Receptor Control Room EAB* LPZ Total Dose (rem TEDE) 1.76 0.530 0.0450 Acceptance Criteria (rem TEDE) 5 25 25

• Worst 2-hr period of the accident duration NMP Unit 1 UFSAR Section XV XV-104 Rev. 25, October 2017 TABLE XV-9 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS GNF2 (Reference 62) A. Plant Parameters Core Thermal Power (MWt) Nominal Appendix-K Vessel Steam Output (lbm/hr) Vessel Steam Dome Pressure (psia) Maximum Recirculation Line Break Area (ft2) Initial MCPR Initial Water Level 1887 (102% of Rated) 7.49*106 (corresponds to 102% rated core power) 1045 7.223 1.45 Nominal Level B. Emergency Core Cooling Systems Parameters Core Spray System System Flow vs. Vessel Pressure Initiating Signals and Setpoints Low Water Level (Downcomer Level) - OR - High Drywell Pressure (psig) Maximum Allowable Delay Time from Initiating Signal to Pump at Rated Speed (sec) (1) This value is added to the maximum degraded voltage time delay in TS Table 3.6.2i for a degraded grid voltage coincident with a LOCA (section XV-C.2.2.5). See Table XV-9b 7.23 ft above TAF 3.606 35(1)

NMP Unit 1 UFSAR Section XV XV-105 Rev. 25, October 2017 TABLE XV-9 (Cont'd.) GNF2 (Reference 62) Injection Valve Stroke Time (sec) Pressure Permissive at Which Injection Valve Opens (psid) B. Emergency Core Cooling Systems Parameters ADS Total Number of Valves in System Total Number of Valves Assumed in Analysis Minimum Flow Capacity of 3 Valves (lbm/hr) at Vessel Pressure (psig) Initiating Signals Low Water Level (Inside Shroud Level) - AND - High Drywell Pressure (psia) Time Delay After Initiating Signals (sec) 23 349.2 6 3 1,618,000 1117 4.24 ft above TAF 3.606 125 NMP Unit 1 UFSAR Section XV XV-106 Rev. 25, October 2017 TABLE XV-9a THIS TABLE HAS BEEN DELETED NMP Unit 1 UFSAR Section XV XV-107 Rev. 25, October 2017 TABLE XV-9b CORE SPRAY SYSTEM FLOW PERFORMANCE ASSUMED IN THE LOCA ANALYSIS Reactor Pressure Core Spray Flow Core Spray Flow (psig)(1) (gpm)(2) (gpm)(3) 335 0 --- 300 732 --- 250 1801 --- 225 2230 0 200 2650 587 150 3270 1550 100 3770 2447 40.3 4260 3284 15.3 4440 3562 0.0 4540 3718 (1) This pressure assumes that the containment pressure is atmospheric. The flow shown will be delivered when this differential pressure between the vessel and the containment exists. (2) One core spray pump and one topping pump. (3) One core spray pump without topping pumps. NMP Unit 1 UFSAR Section XV XV-108 Rev. 25, October 2017 TABLE XV-10 NINE MILE POINT UNIT 1 ECCS SINGLE VALVE FAILURE ANALYSIS System Valve(s) Position for Normal Plant Operation Consequences of Valve Failure Assumed Together with Design Basis LOCA Closed Opened Core Spray Suction Isolation Test Injection(s) (Inside) Check Check (For pressure relief) X X X X X Negate use of one core spray pump and topping pump in same line. No consequences - power removed from valve for normal operation. No consequences - alternate parallel line available. No consequences - assumption of valve failure not required (SECY-77-439). No consequences - assumption of valve failure not required (SECY-77-439). ADS X Vessel depressurizes faster; therefore, the LPCS will be activated earlier and consequences are less severe. NMP Unit 1 UFSAR Section XV XV-109 Rev. 25, October 2017 TABLE XV-11 SINGLE FAILURES CONSIDERED IN LOCA ANALYSIS Single Break Location Failure Available Systems Recirculation Line 1 EC 2 CS2+ 1 EC* + 3 ADS Valves 1 ADS 2 CS2+ 2 EC + 3 ADS Valves** 1 DG 2 CS1+ 2 EC + 3 ADS Valves Feedwater & Main 1 EC 2 CS2+ 1 EC + 3 ADS Valves Steam Lines 1 ADS 2 CS2+ 2 EC + 3 ADS Valves** 1 DG 2 CS1+ 2 EC + 3 ADS Valves Core Spray Line 1 EC 1 CS2+ 1 EC + 3 ADS Valves 1 ADS 1 CS2+ 2 EC + 3 ADS Valves** 1 DG 1 CS1+ 2 EC + 3 ADS Valves EC = Emergency Condenser CS1 = Core Spray (1 Core Spray Pump + 1 Core Spray Topping Pump Feeding 1 Sparger) CS2 = Core Spray (2 Core Spray Pumps + 2 Core Spray Topping Pumps Feeding 1 Sparger) ADS = Automatic Depressurization System DG = Diesel Generator

• The operable EC system is assumed to be attached to the broken recirculation loop. ** Failure of one ADS valve automatically switches the ADS function to a backup set of 3 ADS valves; there will always be 3 ADS valves available.

NMP Unit 1 UFSAR Section XV XV-110 Rev. 25, October 2017 Table XV-12 TRACG-LOCA Licensing Results for Nine Mile Point 1 Parameter 1. Fuel Type 2. Limiting Break 3. Limiting Single Failure 4. Peak Cladding Temperature 5. Maximum Local Oxidation 6. Core-Wide Metal-Water Reaction 7. Coolable Geometry 8. Long-Term Cooling Value GNF2 Recirculation discharge 200% split break 1 Diesel Generator 2105 °F < 12.50% < 1.00% Items 4 and 5 1 core spray available Acceptance Criteria N/A N/A N/A 2200 °F 17%* 1% PCT 2200°F and Maximum Local Oxidation 17% Core temperature acceptably low and long-term decay heat removed

• In compliance with the TRACG-LOCA Safety Evaluation Report (Reference 63), the maximum local oxidation is limited to 13%

NMP Unit 1 UFSAR Section XV XV-111 Rev. 25, October 2017 TABLE XV-13 thru TABLE XV-21a TABLES XV-13 THRU XV-21a HAVE BEEN DELETED NMP Unit 1 UFSAR Section XV XV-112 Rev. 25, October 2017 TABLE XV-22 ACTIVITY RELEASED TO THE REACTOR BUILDING FOLLOWING THE FHA (curies) Noble Gases Halogens Isotope Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m Xe-133m Xe-133 Xe-135 Total Activity 1.06E+02 5.02E+02 1.79E-02 3.34E+01 1.93E+02 9.45E+02 3.25E+04 7.83E+03 4.21E+04 Isotope I-131 (organic) I-132 (organic) I-133 (organic) I-135 (organic) I-131 (elemental) I-132 (elemental) I-133 (elemental) I-135 (elemental) Total Activity 3.83E+01 3.07E+01 2.37E+01 4.00E+00 9.52E+01 7.63E+01 5.89E+01 9.93E+00 3.37E+02 NMP Unit 1 UFSAR Section XV XV-113 Rev. 25, October 2017 TABLE XV-23 UNIFORM UNFILTERED STACK DISCHARGE RATES FROM 0 TO 2 HR AFTER THE FHA (curies/sec) Noble Gases Halogens Isotope Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m Xe-133m Xe-133 Xe-135 Total Activity 1.47E-02 6.97E-02 2.49E-06 4.64E-03 2.69E-02 1.31E-01 4.51E+00 1.09E+00 5.85E+00 Isotope I-131 (organic) I-132 (organic) I-133 (organic) I-135 (organic) I-131 (elemental) I-132 (elemental) I-133 (elemental) I-135 (elemental) Total Activity 5.33E-03 4.27E-03 3.29E-03 5.56E-04 1.32E-02 1.06E-02 8.18E-03 1.38E-03 4.68E-02 NMP Unit 1 UFSAR Section XV XV-114 Rev. 25, October 2017 TABLE XV-24 FUEL HANDLING ACCIDENT DOSES Total Dose Acceptance Criteria Receptor (rem TEDE) (rem TEDE) Control Room 0.847 5 EAB* 0.447 6.3 LPZ 0.0384 6.3

• Worst 2-hr period of the accident duration NMP Unit 1 UFSAR Section XV XV-115 Rev. 25, October 2017 TABLE XV-25 FHA ANALYSIS INPUTS AND ASSUMPTIONS Parameter Value Fuel Failure 2.01 assemblies out of 532 Fuel Decay (Recently Irradiated Fuel) 24 hr Reactor Power Level Analyzed Thermal Power 1887 MWt Peaking Factor 1.79 Gap Fractions I-131 8% Kr-85 10% Noble Gases 5% Halogens 5% Alkali Metals 12% Iodine Speciation Organic 0.15% Elemental 99.85% Cesium Iodide (Particulate) 0% Pool Decontamination Factors Minimum Water Depth 22.75 ft Overall Iodine 191 Elemental Iodine 268 Alkali Metals Infinite Noble Gases 1 Release Duration Instantaneous Control Room Intake Flow Infinite Exchange Rate Control Room Volume 135,000 ft3 Filtration None NMP Unit 1 UFSAR Section XV XV-116 Rev. 25, October 2017 TABLE XV- Parameter Value Atmospheric Dispersion, /Q Turbine/Condenser Release (Ground) EAB 1.90 x 10-4 sec/m3 LPZ 1.63 x 10-5 sec/m3 CR 4.82 x 10-4 sec/m3 Dose Conversion Factors FGR 11 & 12 NMP Unit 1 UFSAR Section XV XV-117 Rev. 25, October 2017 TABLE XV-26 CRD ACCIDENT ANALYSIS INPUTS AND ASSUMPTIONS Parameter Value Fuel Failure Fuel Melt Fraction 0 Cladding Failure Fraction 2.577% of core Reactor Power Level Analyzed Thermal Power 1887 MWt Peaking Factor 1.8 Fuel Release Fraction to Coolant Gap Iodine 10% Noble Gas 10% Alkali Metals 12% Halogens 5% Tellurium Group 0 Reactor Turbine/ Iodine Speciation Coolant Condenser Organic 0.15% 3% Elemental 4.85% 97% Cesium Iodide (Particulate) 95% 0 Reaches the Available Condenser Release Fractions Condenser for Release Iodine 10% 10% Noble Gas 100% 100% Alkali Metals 1% 1% Halogens 1% 1% Tellurium Group 1% 1% Leakage Parameters Main Condenser Volume 50,000 ft3 Main Condenser Leak Rate 1%/day Mechanical Vacuum Pump Exhaust Rate 280,000 lbm/hr Release Duration 24 hr NMP Unit 1 UFSAR Section XV XV-118 Rev. 25, October 2017 TABLE XV-) Parameter Value Control Room Intake Flow Infinite Exchange Rate Control Room Volume 135,000 ft3 Filtration None Atmospheric Dispersion, /Q Turbine/Condenser Release (Ground) EAB 1.90 x 10-4 sec/m3 LPZ 1.63 x 10-5 sec/m3 CR 1.03 x 10-3 sec/m3 Atmospheric Dispersion, /Q Mechanical Vacuum Pump Release (Elevated) EAB 5.98 x 10-5 sec/m3 LPZ 2.12 x 10-5 sec/m3 CR 2.27 x 10-4 sec/m3 NMP Unit 1 UFSAR Section XV XV-119 Rev. 25, October 2017 TABLE XV-27 CRDA NOBLE GAS RELEASE Time After Stack Discharge Rate* Accident (Curies/Sec) 3.6 sec 6.60 x 10-3 30 min 4.28 1 hr 6.70 1.5 hr 6.98 2 hr 6.85 4 hr 5.97 8 hr 4.55 12 hr 3.42 1 day 2.38
• Average rate over preceding time interval.

NMP Unit 1 UFSAR Section XV XV-120 Rev. 25, October 2017 TABLE XV-28 CRDA HALOGEN RELEASE Time After Stack Discharge Rate* Accident (Curies/Sec) 3.6 sec 6.39 x 10-5 30 min 6.43 x 10-2 1 hr 1.13 x 10-1 1.5 hr 1.19 x 10-1 2 hr 1.15 x 10-1 4 hr 9.45 x 10-2 8 hr 6.62 x 10-2 12 hr 4.61 x 10-2 1 day 2.79 x 10-2

• Average rate over preceding time interval.

NMP Unit 1 UFSAR Section XV XV-121 Rev. 25, October 2017 TABLE XV-29 CONTROL ROD DROP ACCIDENT DOSES Case 1 Case 2 Acceptance Total Dose Total Dose Criteria Receptor (rem TEDE) (rem TEDE) (rem TEDE) Control Room 0.610 1.60 5 EAB* 0.630 0.340 6.3 LPZ 0.0540 0.210 6.3

• Worst 2-hr period of the accident duration.

NMP Unit 1 UFSAR Section XV XV-122 Rev. 25, October 2017 TABLE XV-29a WETTING OF FUEL CLADDING BY CORE SPRAY Time Condition 0-1800 sec Heat transfer by radiation. 1800 Wetting of fuel cladding begins. 3600 Wetting complete. NMP Unit 1 UFSAR Section XV XV-123 Rev. 25, October 2017 TABLE XV-29b POST-LOCA AIRBORNE DRYWELL FISSION PRODUCT INVENTORY (curies) Cesium and Other Time (hr) Noble Gases Halogens Rubidium Solids* 0.417 9.46E+06 2.08E+06 1.38E+05 1.97E+04 0.917 6.05E+07 2.95E+06 1.73E+05 4.73E+05 2.917 9.96E+07 2.12E+05 1.14E+04 2.96E+04 8.033 8.10E+07 4.80E+04 1.10E+01 2.55E+01 24 6.17E+07 2.70E+04 2.49E-03 4.92E-03 96 3.65E+07 1.04E+04 2.36E-03 3.79E-03 240 1.57E+07 5.26E+03 2.13E-03 2.60E-03 720 1.26E+06 7.69E+02 1.61E-03 1.30E-03

• Except particulate iodine NMP Unit 1 UFSAR Section XV XV-124 Rev. 25, October 2017 TABLE XV-29c POST-LOCA REACTOR BUILDING FISSION PRODUCT INVENTORY (curies) Cesium and Other Time (hr) Noble Gases Halogens Rubidium Solids* 0.417 1.21E-01 1.71E+00 0.00E+00 0.00E+00 0.917 4.03E-01 2.62E+00 0.00E+00 0.00E+00 2.917 1.52E+00 1.86E+00 0.00E+00 0.00E+00 8.033 1.81E+05 1.98E+04 8.92E-02 2.06E-01 24 6.66E+05 5.29E+04 2.33E-02 4.61E-02 96 2.68E+05 2.29E+04 4.70E-05 7.57E-05 240 1.12E+05 9.16E+03 1.23E-05 1.50E-05 720 7.81E+03 6.23E+02 9.28E-06 7.51E-06
• Except particulate iodine.

NMP Unit 1 UFSAR Section XV XV-125 Rev. 25, October 2017 TABLE XV-29d POST-LOCA DISCHARGE RATES (curies/sec) Cesium and Other Time (hr) Noble Gases Halogens Rubidium Solids* Filtered Stack Release 0-0.417 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.417-0.917 4.81E-02 3.89E-02 0.00E+00 0.00E+00 0.917-2.917 8.51E-01 1.94E-01 0.00E+00 0.00E+00 2.917-8.033 2.62E+00 1.20E-01 3.22E-08 7.56E-08 8.033-24 1.29E+01 5.96E-02 6.54E-08 1.40E-07 24-96 9.34E+00 4.58E-02 4.49E-09 8.53E-09 96-240 4.50E+00 1.89E-02 2.14E-11 1.26E-11 240-720 9.41E-01 4.09E-03 1.27E-11 1.15E-11 Unfiltered Ground-Level Release 0-0.417 9.83E-01 7.14E-01 2.01E-02 2.62E-03 0.417-0.917 6.32E+00 1.23E+00 2.83E-02 5.70E-02 0.917-2.917 2.97E+01 6.70E-01 2.64E-02 6.83E-02 2.917-8.033 2.25E+01 6.69E-02 1.53E-03 3.42E-03 8.033-24 6.00E+00 4.13E-03 9.96E-05 2.00E-04 24-96 2.09E+00 7.06E-04 4.23E-07 1.13E-09 96-240 1.12E+00 3.26E-04 0.00E+00 0.00E+00 240-720 2.41E-01 1.04E-04 0.00E+00 0.00E+00

• Except particulate iodine NMP Unit 1 UFSAR Section XV XV-126 Rev. 25, October 2017 TABLE XV-30 CORE FISSION PRODUCT INVENTORY Nuclide Ci/MWt Nuclide Ci/MWt Nuclide Ci/MWt Kr-83m 3.27E+03 Ru-106 1.76E+04 Cs-134 7.29E+03 Kr-85 3.93E+02 Rh-105 2.84E+04 Cs-136 2.28E+03 Kr-85m 6.82E+03 Sb-127 3.01E+03 Cs-137 4.35E+03 Kr-87 1.30E+04 Sb-129 8.91E+03 Ba-137m 4.12E+03 Kr-88 1.83E+04 Te-127 3.00E+03 Ba-139 4.89E+04 Kr-89 2.22E+04 Te-127m 4.05E+02 Ba-140 4.71E+04 Rb-86 7.29E+01 Te-129 8.76E+03 La-140 5.12E+04 Sr-89 2.45E+04 Te-129m 1.30E+03 La-141 4.45E+04 Sr-90 3.14E+03 Te-131m 3.97E+03 La-142 4.29E+04 Sr-91 3.10E+04 Te-132 3.85E+04 Ce-141 4.47E+04 Sr-92 3.38E+04 I-131 2.71E+04 Ce-143 4.11E+04 Y-90 3.24E+03 I-132 3.92E+04 Ce-144 3.70E+04 Y-91 3.18E+04 I-133 5.51E+04 Pr-143 3.97E+04 Y-92 3.40E+04 I-134 6.03E+04 Nd-147 1.80E+04 Y-93 3.96E+04 I-135 5.16E+04 Np-239 5.78E+05 Zr-95 4.46E+04 Xe-131m 3.04E+02 Pu-238 1.45E+02 Zr-97 4.51E+04 Xe-133 5.27E+04 Pu-239 1.34E+01 Nb-95 4.48E+04 Xe-133m 1.63E+03 Pu-240 1.89E+01 Mo-99 5.13E+04 Xe-135 1.91E+04 Pu-241 5.49E+03 Tc-99m 4.49E+04 Xe-135m 1.09E+04 Am-241 7.48E+00 Ru-103 4.29E+04 Xe-137 4.80E+04 Cm-242 1.85E+03 Ru-105 3.01E+04 Xe-138 4.50E+04 Cm-244 1.23E+02 NMP Unit 1 UFSAR Section XV XV-127 Rev. 25, October 2017 TABLE XV-31 LOCA ANALYSIS INPUTS AND ASSUMPTIONS Parameter Value Reactor Power Level 1887 MWt Early Fission Product Release Fractions Gap Phase In-Vessel Halogens 5% 25% Noble Gases 5% 95% Alkali Metals 5% 20% Tellurium Group 0 5% Ba, Sr 0 2% Noble Metals 0 0.25% Cerium Group 0 0.05% Lanthanides 0 0.02% Release Timing Onset Duration Gap Phase 2 min 0.5 hr Early In-Vessel 0.5 hr 1.5 hr Suppression Iodine Speciation Drywell Pool Organic 1.5% 3% Elemental 4.85% 97% Cesium Iodide (Particulate) 95% 0 Release of Activity to Suppression 30% of core iodine Pool inventory ESF Leakage Leak Rate to Reactor Building 1,200 gph Flashing Fraction 10% Containment Leakage (0 to 24 hr) Primary Containment Leak Rate 1.5 weight % per day Non-MSIV Reactor Building Bypass 91 scfh MSIV Reactor Building Bypass 100 scfh total; 50 scfh max per line Containment Leakage (24 to 720 hours) 50% reduction Reactor Building Drawdown 6 hr NMP Unit 1 UFSAR Section XV XV-128 Rev. 25, October 2017 TABLE XV-d.) Parameter Value Volumes Drywell Airspace (Min) 180,000 ft3 Wetwell Vapor Space (Min) 120,000 ft3 Suppression Pool (Min) 79,700 ft3 Reactor Building Free Volume 2,100,000 ft3 Reactor Building Holdup Volume 3.01E+10 cc Control Room 135,000 ft3 Flow Rates RBEVS 1,600 cfm Control Room Intake 2,025 cfm Assumed CR Unfiltered In-Leakage 100 cfm Filtration CR RBEVS Organic 90% 90% Elemental 95% 95% Particulates 95% 95% Isolation Time <2 min <2 min Fission Product Removal Inputs Drywell Spray Flow Rate 6,383 gpm Drywell Accident Conditions 35 psig, 281°F Steam Line T/H Conditions 1,050 psia saturated conditions Main Steam Line Volume 82.4 ft3 (inboard to outboard MSIV; each line) Breathing Rates CR Offsite 0 to 8 hr 3.5E-4 m3/s 3.5E-4 m3/s 8 to 24 hr 3.5E-4 m3/s 1.8E-4 m3/s 24 to 720 hr 3.5E-4 m3/s 2.3E-4 m3/s CR Occupancy Factors 0 to 24 hr 1.0 24 to 96 hr 0.6 96 to 720 hr 0.4 NMP Unit 1 UFSAR Section XV XV-129 Rev. 25, October 2017 TABLE XV- Parameter Value Dose Conversion Factors FGR 11 & 12 Atmospheric Dispersion, /Q Tables XV-35a, XV-35b, XV-35c and XV-35d NMP Unit 1 UFSAR Section XV XV-130 Rev. 25, October 2017 TABLE XV-32 LOSS OF COOLANT ACCIDENT DOSES Total Dose Acceptance Criteria Receptor (rem TEDE) (rem TEDE) Control Room 4.81 5 EAB* 9.02 25 LPZ 1.60 25
• Worst 2-hr period of the accident duration NMP Unit 1 UFSAR Section XV XV-131 Rev. 25, October 2017 TABLE XV-32a SIGNIFICANT INPUT PARAMETERS TO THE DBR CONTAINMENT SUPPRESSION CHAMBER HEATUP ANALYSIS Plant Parameters Core Thermal Power (MWt) 1887 (102% Rated) Initial Dome Pressure (psia) 1045 Initial Drywell and Wetwell Pressure (psia) 14.7 Maximum Recirculation Line Break Area (ft2) 5.45 Initial Torus Water Level (ft of downcomer submergence) 3.5 Initial Torus Water Temperature (°F)85 Maximum Raw (Lake) Water Temperature (°F) 84 Initial Suppression Chamber Pool Volume (ft3) 79,800 Initial Wetwell Airspace Free Volume (ft3) 125,000 Initial Wetwell Airspace Temperature (°F) 105 Drywell Free Volume (ft3) 180,000 Initial Drywell Temperature (°F) 150 Emergency Core Cooling System Parameters Core Spray System Single Failure See Section XV-C.2.0 Flow vs. Reactor Pressure See Table XV-9a NMP Unit 1 UFSAR Section XV XV-132 Rev. 25, October 2017 Table XV-32a (Cont'd.) Containment Spray System Assumptions Loss of Offsite Power Single Failure Loss of One Emergency (See Section VII) Diesel Number of Containment Spray Pumps: Assumed to Auto Start 2 Available for Heat Removal 1 Available for Spray Assuming No Heat Removal 2 Number of Containment Spray Raw Water Pumps Manually Started 1 Number of Containment Spray Heat Exchangers 1 Time of Heat Exchanger Activation (min) 15 Containment Spray Pump Flow 1-Pump Operation Spray Mode (gpm) 3600 2-Pump Operation Spray Mode (gpm) 3000 per pump Torus Cooling Mode Flow (gpm) 2800 maximum Containment Spray Raw Water Pump Flow (gpm) 3000 Containment Spray Heat Exchanger K-Value in Spray Mode (Btu/sec °F) 256 Containment Spray Heat Exchanger K-Value in Torus Cooling Mode (Btu/sec °F) 241 NMP Unit 1 UFSAR Section XV XV-133 Rev. 25, October 2017 TABLE XV-33 THIS TABLE HAS BEEN DELETED NMP Unit 1 UFSAR Section XV XV-134 Rev. 25, October 2017 TABLE XV-34 THIS TABLE HAS BEEN DELETED NMP Unit 1 UFSAR Section XV XV-135 Rev. 25, October 2017 TABLE XV-34a RELEASE/INTAKE ELEVATIONS Elevation Elevation Point of Interest (ft) (m) Main Stack 350 106.7 Reactor Building Blowout Panel (relative to bottom of panel) 78.9 24 Turbine Building Blowout Panel (relative to bottom of panel) 72.4 22.1 Control Room Intake (height equal to roof elevation) 72 29.95 Technical Support Center 21 6.4 NMP Unit 1 UFSAR Section XV XV-136 Rev. 25, October 2017 TABLE XV-34b RELEASE/INTAKE DISTANCE AND DIRECTIONS Release/Intake Horizontal Distance (ft) Horizontal Distance (m) Sector Bearing Relative to True North Unit 1 Reactor Building Blowout Panel (from midpoint of panel)/Unit 1 Control Room Intake 340 103.6 149° SSE Unit 1 Turbine Building Blowout Panel (from midpoint of panel)/Unit 1 Control Room Intake 236 71.9 117° ESE Unit 1 Main Stack/Unit 1 Control Room Intake 400 121.9 166° SSE Unit 1 Reactor Building Blowout Panel (from midpoint of panel)/Unit 1 Technical Support Center 343 104.5 86° ESE Unit 1 Turbine Building Blowout Panel (from midpoint of panel)/Unit 1 Technical Suport Center 328 100.0 86° E Unit 1 Main Stack/Unit 1 Technical Support Center 330 100.6 140° SE NMP Unit 1 UFSAR Section XV XV-137 Rev. 25, October 2017 TABLE XV-35 THIS TABLE HAS BEEN DELETED NMP Unit 1 UFSAR Section XV XV-138 Rev. 25, October 2017 TABLE XV-35a /Q VALUES FOR THE CONTROL ROOM Release Point 3) 0-2 hr 2-8 hr 8-24 hr 1-4 days 4-30 days Unit 1 Reactor Building Blowout Panel 4.82E-04 2.61E-04 9.25E-05 6.70E-05 4.93E-05 Unit 1 Turbine Building Blowout Panel 1.03E-03 5.85E-04 2.07E-04 1.75E-04 1.52E-04 Unit 1 Main Stack 2.27E-04 1.26E-04 4.30E-05 3.58E-05 2.59E-05 NMP Unit 1 UFSAR Section XV XV-139 Rev. 25, October 2017 TABLE XV-35b /Q VALUES FOR THE TECHNICAL SUPPORT CENTER Release Point 3) 0-2 hr 2-8 hr 8-24 hr 1-4 days 4-30 days Unit 1 Reactor Building Blowout Panel 7.09E-04 5.60E-04 2.345E-04 1.71E-04 1.41E-04 Unit 1 Turbine Building Blowout Panel 5.91E-04 4.26E-04 1.63E-04 1.35E-04 1.16E-04 Unit 1 Main Stack 3.47E-04 2.42E-04 8.22E-05 6.06E-05 5.00E-05 NMP Unit 1 UFSAR Section XV XV-140 Rev. 25, October 2017 TABLE XV-35c OFFSITE /Q VALUES FOR GROUND-LEVEL RELEASES Boundary 3) 0-2 hr 0-8 hr 8-24 hr 1-4 days 4-30 days EAB 1.90E-04 --- --- --- --- LPZ --- 1.63E-05 1.10E-05 4.67E-06 1.67E-06 NMP Unit 1 UFSAR Section XV XV-141 Rev. 25, October 2017 TABLE XV-35d OFFSITE /Q VALUES FOR ELEVATED RELEASES Boundary 3) 0-2 hr 0-8 hr 8-24 hr 1-4 days 4-30 days EAB 5.98E-05 --- --- --- --- LPZ --- 2.12E-05 8.40E-07 3.45E-07 1.11E-07 NMP Unit 1 UFSAR Section XV XV-142 Rev. 25, October 2017 TABLE XV-36 REACTOR BUILDING LEAKAGE PATHS Joint Length Area Walls 2@ 5,147 ft 2@ 9,370 ft2 Walls 2@ 3,778 ft 2@ 6,850 ft2 Roof 550 ft --- Total 18,400 ft 32,440 ft2 STATION TRANSIENT DIAGRAM -------*r*-------------------------------------------------------------------*-*i-*--------------------------------------------------------------. ! BALANCE OF STATION ---------*-*--*-*---*-REACTOR MAIN STEAM LINE -FEEOWATER ,,,,.----------, / \ @ rzp (i)f1A\112\ ( _ I I I I I I I : I I // ', / ' / \ / \ I \ I \ I I I I ! © \ I I I l ORYWELL I I I I \ I \ I \\ 0@@ / \ I \ / \ / ',, _//// ................. _ ----------REACTOR SCRAM Loop NO SCRAM 4. Control Rod W1thdr-owol Error 7. Rec1r-culeit1on Pump T l"lps B. Rec1r-culertJ.on Pump Stall CJ. Flow Contt-oller-Mislfunct1on Cincr-eos1n9 Flow) H!l. Flow Controller-Molfunct1on COecr-eosing Flow) 11. lnodver-tent Aotuot1on of One SolennoLd Aotuo-ted Relief Volvo REACTOR SCRAM 5. Mam Steam LLne IsoJ.i:it1on Volve Closu.-e 13. Feedw111ter Controller Mel function CZe.-o Flow) TURBINE ---CONDENSER 3. Fe.,dwnter Controlll'r Malfunction CMalnmum Flow] I. Turbine Trip -Fo1lure of Bypesl! System 14. Tu.-b1ne Tr-1p -P!!lt-"tlol Byposs -(nter-med1ote Power 15. Turbine Trip -Por-t1ol Syposs -Stretch Powei-NO SCRAM 16. Inodver-tent Actuot1on Byposs Volve 21. Pressure Regulotor-Molfunct1onve 17. One Feedwoter-Pump Tr-Lp end Restart 2. Loss of 100°F Feedweter-Hee.ting TURBINE GENERATOR -CONDENSER -ELECTRICAL 345KY 115KY 4161ilV 4160V )@a_ -1.i 4160V 4L6e!V 6"0V REACTOR SCRAM 416121V 18. Loss of Moin Condenser-Vocuum !CJ. Loss of Electr1cel Loeid 20. Loss of Aux1hor-y Po .. er-22. Instrument A1r-Fo1lur-e 23. O.C, Power Interruptions NO SCRAM 24. F edur-e of One 01ei1el Generetor-to Ster-t 25. Power-Bus Loss of Volt.eige FIG. XV-I UFSAR Rev1s1on 17 October 2001 Nine Mile Point Unit 1 FSAR FIGURE XV-2 THIS FIGURE HAS BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 150. 0 Cl 100. 0 < a: IL c .... z: "' -50. 0 o. 0 ISO. 0 C so. 0 o. 0 o.o ---0.0 PLANT RESPONSE TO LOSS OF 100°F FEEDWATER HEATING l NEU RON FLUX 2 AVE SURFACE HEAT FLUX 3 COR INLET FLO\/ I 00. 0 TIHE ISECONOSJ I LEV LCINCH-REF-SEP-SKRT) 2 VES >EL STEAMFLO\I 3 TUR HNE STEAMFLO\I
• r--1ua .... r-n -. . . . . . I 00. 0 TIHE !SECONDS] 200.0 200.0 ... en .. 150.0 0 50.0 0.0 I. 0 ffi o. 0 c u ,_ ... ... !;! -J. 0 -2.0 .. o. 0 o. 0 ----l VES >EL PRESS RISECPSIJ 2 REL EF VALVE FLO\/ 3 BYP ,55 VALVE FLO\/ . ,., . :;i 1 *. " --I 00, 0 TltlE (SECONDS J I 00. 0 TlHE C SECONDS) --.. .., -FIGURE XV-3 .. -200. 0 200.0 UflAR Revision 14 (June 1996)

Nine Mile Point Unit 1 FSAR FIGURE XV-4 THRU FIGURE XV-7 FIGURES XV-4 THRU XV-7 HAVE BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 Cl w I-a: a: LL D I-z w Ll a: w a... 125. 75. 25. -25. 0. STARTUP OF COLD RECIRCULATION LOOP PARTIAL POWER 0 NEUlAON FLUX. ¢ SURFACE HEAT + RE JRCULAT.ION X VE SEL S1EA1'1 0/0 LUX. FLOIJ. LOIJ, I FEEOIJAlER FLO , 0/0 0/0 0/0 0/0 llll1'X9 Y. 8. 12. 16. TIME (SEC. J STARTUP OF COLO 0 VESSEL PRES C ANGE; PST ¢ ACTUAL LEVEL HANCE. TN. TIME FIGURE XV-8 UFSAR Revision 14 WuM 1991) (-, c IJ,J I-a: a: u.. c I-z LLJ u a: LLJ a.. too. so. RECIRCULATION PUMP TRIPS (1 PUMP) e. TIME t2. f SEC. l TRIP CF CINE TRJP CF CINE ts. tao. ----e. TIME t2. fSEC.l t 1 VESSEL PRES KGE.PSJ uRCT, Sl\JRT HERSJCIN,JN 'J CCIAE SUBCCIOLJ G. envLB. ti. FIGURE XV-9 Uf8AR Reviaien 14 (Jwle 1 ... ( c Lr.J t-ct a: u... IO t-2 Lr.J u a: Lr.J 0... RECIRCULATION PUMP TRIPS (5 PUMPS) TftJP fl f'JVE \SO. J C. PUltP 11 "EUTAOH nux u Sl#ACE HCAf LUX '1 AtCJACUl.ATIO f'LOW u VESSEL STEAH LOW ') FttOWATEA no 100. so. 4 8. \2. TJME ISEC. l fAJP fl f' JVt TJME \2, r SEC. l 11VtSSEL PAtS HC.t.PSJ 12Acr. St\JRT Ht1'SJOH.JH 11 SUBCOOLJ Ci, 8T\J/L8. \6. *FIGURE XV-10 UFSAR Revision 14 (.lune 1996) ( c w I-c: a: IJ.. 0 I-z w u a: w CL.. 1SO. '00. so. RECIRCULATION PUMP STALL SEJ WRE OF ONE RECJFIC ** PUHP 18 g a. TJME 1 i NEUTRON FLUX u SURf'RCE HEAT 12. rSEC. 1 u RECIFICULATJO **VESSEL STEAH n FEEDMATER fl SEIWFIE Of ONE LUX FLOM LOW 16. ) . TIME fSEC. \ FIGURE XV-11 UFSAR Revision 14 WuM 1196) ( 100. so. .. o. FLOW CONTROLLER MALFUNCTION (INCREASED FLOW) 8. TIME * .. ** .. . i..---. "' .. .. .. .... ... .. .. ... I I I I I I I I I e. TIME LSEC..l .. CSEC. 1 12.

• NEUTRON FLUX 1 SUAFRCE HERT LUX 1 AECIRCULATIC:
• VESSEL STEAM *LJW I fEEOWRTEfi FL " 16. ' I
• VESSEL PRES a ACT. SKIRT S 3MEF.SION. IN 12. 16. FIGUM XV-12 UfSAR Revi-slon 14 (June 1H8)

( ---\ \SO. c UJ \00. .... a: ct: 0 .... z UJ u so. ct: UJ ./ FLOW CONTROLLER MALFUNCTION DECREASING FLOW "* a. TIME a. TIME 12. f SEC.) 12. f SEC. l I l NEUTRON f'LUX H Mf'RCE HERT u RECJRCULRT JO u VESSEL STERH FEEDWRTER f'L f'LOW CONT. f'AJ 1 t VESSEL PRES u ACT. SKJRT 16. 16. RGURE XV-13

• UFSAR Revision 14 (June 1991)

INADVERTENT ACTUATION OF ONE SOLENOID RELIEF VALVE 150, c LL.I 100. a: a: LL. IC 1-z LL.I u so. a: LL.I .,,. . .. .. .. .. -.. .. ..

• I I I I I I a'o. 100. so.
• a. . .. -I I I I I * . . . . * . . . . . I o 2. TIME ' . . . . . 2. ". TIME ONE AELJEf' RCCJD OPENlNG 1 !SO I 1 NEUTRON f'LUX u SWACE HEAT *l . u YESSE. ST£8H , . 4 . 2 l . . . 3 6. 8. rSEC.l ONE AELJEf' YALY RCCJD OPENING 1 11 YESStL .HANGE.PSJ nACT.Sl'\JRT SU JN. . . . 2 . I 6. 8. (SEC. l FIGURE XV-14 UFIAR Revision 14 (.June 1991)

Nine Mile Point Unit 1 FSAR FIGURE XV-15 THRU FIGURE XV-16 FIGURES XV-15 THRU XV-16 HAVE BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 ( FEEDWATER CONTROLLER MALFUNCTION -ZERO FLOW LOSS OF FEEOWAT

• 1 NEUTRON FLUX u SURFACE HEAT u RECJRCULATJO **VESSEL STEAH LOW C5FEEOWATER FLO fa CC a: u.. I!:) 1-z LU U 50, 4 a: LU Q_ 10. TIME 15. fSEC. l 20. LOSS OF FEEOWAT 10. TIME ll. f SEC. l *1VESSEL PRES ANGE.PSI t2ACT. Sl'\JRT SU HERSION,JN 20. FtQUftE XV-17 UflAR Revision 14 (June 1-)

TURBINE TRIP WITH PARTIAL BYPASS INTERMEDIATE POWER ( MBJHE TRJP tSO. 11 2 u REC J RCULA Tl 0 **VESSEL STEAH cs FEEOWRTER FL Cl LU 100. I-a: cc. u.. c I-2 LU u so. cc. LU 2 4 1 o.o. e. 12. 16. TIME <SEC.) TURBJNE TRIP '200. TAJ P SCAAH 980H lff 1 1 VESSEL PRES1E '"HRNGE.PSJ uACT.SKJRT SU iERSION JH. 100. . . . . . I* o. . lo ""' . . lo . .. -i-* .. I I ... e. \2. 16. Tl ME fSEC.) FIGURE XV-18 UFIAR Revision 14 (June 1996)

l. / TURBINE TRIP WITH PARTIAL BYPASS Q l&J a: iC CJ z l&J a.. TIME TIME a. B. 12. CSEC.l 12. CSEC.l FIGURE XV-19 UFSAR Revision 14 (June 199G)

., .-..; INADVERTENT ACTUATION OF ONE BYPASS VALVE ( ONE BTPRSS 1SO. RCCJD OPENJNG 1 !SO 1 1 NEUTRON FLUX u SUAFRCE HERT . *' FLOW s . . *

• 4 I 'I *u
• 4 --.J *-* ,..,.. r .... 2 1 CJ 100. LI.I . . . lo. . . l ..... ct a: IJ... l!::> ..... . :z LI.I u so. a: Lt.J .. 0.... ... -... .. .. * * *
• I * * * * "* 6. 8. TIME lSEC. l ONE BTPRSS VALV \ 00. ACCID CPENJNG \ iSO I I VESSEL PRESS '"HRHGE. PSJ u ACT.SKJAT SUB 1EASJON JN. so. . * . .. . . 2 D. I ... -.. ... .. I * . *
• I o 2. "* 6. 8. TIME (SEC. l AQURE XV-20 UFIAR Revision 14 CJune 1 H6) 1SD. ...... c \00. UJ t-a: a: l.L E) :z UJ u SD. a: .. UJ ... loo ... i-. ... .. ... loo'** so. a. -* * ... -* * *
• ONE FEEDWATER PUMP TRIP AND RESTART * .. . . . . . ' I ' I I ' \0. TIME * . . ' ' '
• I ' I ' ' \0. TIME * *
• 20.
• 20. f"'. .. f SEC. l . I f SEC .* \ ONE f'OW PUtP TR p AHO FIESTMT \8 11 HEUTAOH fl.UX ustJV'ACE HERT '! fl.OW * *. . I s
• ii' r ...... ,,,..n t.n r
• 2 1 * *
• mE P1JA. l"IJMP'" r.a l>P RHO RESTART 18 I I VESSEL PRESS :'HRHGE.P5J 12 SUB JN. . .
• z I FIGURE XV-21 UFSAR Revision 14 (June 1996)

( **., 0 UJ I-0: a: u.. IO I-z UJ u* a: UJ LOSS OF ELECTRICAL LOAD LOSS tJr ELECT L RO 100. so. I 1 NEUTRON f'LUX --'---4---1..---+-----a---i-2 URF'RC£ HEAT LUX II. 8. TIME CSEC. l 12. u CJACULRTJO f'LOW u V SEL STERH LOW cs FE WATER fl ts. 11VESSEL PRES RHGE.PSJ *zRCT. SKJRT HEASJON,JN TIME CSEC. l AGUREXV-22 UFSAJll Revision 14 (June , ... , (-"', c L&J r-a: a: lL E> r-z L&J u a: L&J Q.. 100. so. LOSS OF AUXILIARY POWER e. *12. TIME CSEC. l J

• t£UTRON Fl!JX a SUAfRCE LUX 1 AECIRCULAT£01
• VESSEL LCli' 1 4
• 16.
• VESSEL PRES 1 fl:T. SKIRT S MEFSION. IN e. 12. 16. TIME CSEC. l R*GURE XV-23 UFIAlll Revision 14 (June 1991) u... <( 0 en Q... Q... UJ :E 0::: UJ 1600 1200 1000 ::> I-800 UJ -0::: ::> Q... *ii Ei :S. I t s: ... 600 400 ' .. 200 r ... I CJ) -0 \ '< -\ PRESSURE REGULATOR MALFUNCTION INITIAL PRESSURE REGULATION MALFUNCTION SLOWDOWN TO BYPASS VALVES HOT STANDBY -_.....START OF MIXTURE SLOWDOWN 13 SEC I I I I OF MIXTURE BLOWDOWN 43.5 SEC ," "" ' WATER ' TEMPERATURE SAT Of ......._ ' '"""'-i'--.... i"'oio....... PRESSURE -PS"iA'-"-............... 40 80 120 160 200 240 280 320 360 TIME -SECONDS "'D ::0 m (/I (/I c ::0 m ::0 m Cl c s: -f 0 ::0 3:: > "Tt c z n -f -0 z

-LiJ 0::: 0 -c:..:i LL-0 c.. .....__o I-UJ Vl _o c.. x UJ -MAIN STEAM LI NE BREAK -COO LANT LOSS Cl C-..0 I I I I I (..) I UJ I Vl --J I c ! UJ Vl 0 -I I c:..:i I -I I §I I

• I I I -I I UJ Vl Vl UJ I > I Vl " Vl 0 -I Vl Vl " < ::E I-UJ enc 0-1 Cl Cl --1-1-=> z:CC <U.I -I z: o-occ I-= co (S81 OOOtX) SSVW " ""' "" ' ROUREXV-25 Cl C-..0 ' UFIAR Revision 14 (June 1991) <D -

Nine Mile Point Unit 1 FSAR FIGURE XV-26 THRU FIGURE XV-56 FIGURES XV-26 THRU XV-56 HAVE BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 Nine Mile Point Unit 1 FSAR FIGURE XV-56A THRU FIGURE XV-56C FIGURES XV-56A THRU XV-56C HAVE BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 LL. 0 . 0.... ::a:: LU I-1500 1000 *1* -500 " -f ...,. I

• OI 1"8 ...
• f ... I ...._. / __,,; j I j ) ) j )1 II' 'I " I i.,-t::--'""" i..... LOSS-OF-COOLANT ACCIDENT -WITH CORE SPRAY CLADDING TEMPERATURE / ...-0% ............... J ....,,, .. I I\, 10% I ... loo"" ........... ['I '"""" ... I/ II" ' I" ) 25% I I \. \ ....... --L... ... ...... I v ... 1"11 ""' \ I\ / 50% // v "" ... ......... .... "'-' \ ' ,,,,,,,. ""'"" v ,, If ,, 75% v i...i...""' ,......, " \\ / i.-"""" ""'"" !.-1 ..... / v ' , v .. v \ ""'""' 100% 1 ""'"' "' .. \ ' , CORE SPRAY ACTUATED \ l -+' l\\ I\ la I CLAD WETTIN G 11 STARTS I 10 102 10 3 TIME AFTER ACCIDENT SEC I I 10 4 30 (.!) J 20 a.. !ii Ei ., )( .10 ti I ....
• f ... I -en "" [\ \ ' \-I' I' i--,.._ 10 -LOSS-OF-COOLANT ACCIDENT DRYWELL PRESSURE I I ' / ' \ \ v '--. " \ I\ ""'""' ""'"' -10 2 10 3 TIME AFTER ACCIDENT SEC I I ' -NO CORE SPRAY I"--.. r--... ... -I"' .... CORE SPRAY I I I I 10 4 l' C'.!J en Q_ -i 30 20 J r io s. I . "' 8"* :s ,. ...
• I i 0 -.... II" "" .... v I , I I 1 10 LOSS-OF-COOLANT ACCIDENT SUPPRESSION CHAMBER PRESSURE I ' , / ' )' \ '--.. \ I\' " I', -102 103 TIME AFTER ACCIDENT SEC NO CORE SPRAY r---.... ,....... ..... '""""""' CORE SPRAY. l I I I .** I 104 ..

LL. 0 300 ) .. 200 lJJ I-) =-i 100 l>< << -I !I ... .... f ... I --LOSS-OF-COOLANT ACCIDENT CONTAINMENT TEMPERATURE-WITH CORE.SPRAY DRYWELL l""O ...... I'\ r-.... r-.-,....,.... """' ii---,_,.. ----i--SUPPRESSION CHAMBER J __ .......... 10 102 103 TIME AFTER ACCIDENT SEC -.. ' """""' ...... ,. 104 I I 3000 -LI. 0 -41 L :J .... 2000 IO L QJ a. E QJ J-!iii 1000 Ei i'>< << -. .. UI a* ..... :::s ... .... f ... I -CONTAINMENT DESIGN BASIS CLAD TEMPERATURE RESPONSE:. WITHOUT CORE SPRAY , ) )_ V ( ' J i j I v ........... \oV ' ' I / "' I I 31 / I "/ /v ........ *::: . ; I , .,,-." ,,. .... """ ... -/ ......... 100% *. i:;---*** i--11 v"v _vl 1 10 102 103 1011 Time After Accident (sec) flP c 0 +:: u IU QJ 0:: L == -l1J .... QJ :E ' ) ( \. co N . .... .... 0 >-C'D L a. (/) Q) L 0 u \ \ ' CONTAINMENT DESIGN BASIS METAL -WATER REACTION * ' c N N .... --co .............. FIGURE XV-58 c: ----o---->---C'D L--a. __ (/) f-..... 0 u .. " M 0 ... N 0 ... Q ... Q . .... .... . QC -(J Q) Ill -... c: Q) :2 (J u < L Q) .::::= < Q) E I-UflAll Revision 14 (June 1991) \ \' 0 0 -CONTAINMENT DESIGN BASIS CLAD PERFORATION WITHOUT CORE SPRAY LL 0 0 0 0 M 41 > .8 < . Ill c. E 41 .... -41 :s --------.........._ 0 CIO c: 0 Ill L L 41 0. "O Ill u LL ............. -.............. 0 ID -----.......... ............_ 0 =-...... ...._ ................... 0 N " " \ -.... ' ... " "' \ \ , RQUIEXV-59 =-0 .... M 0 .... N 0 0 --CJ 41 Ill -... c: 41 "O *u CJ < L 41 ¢: < 41 E .... UFSAR Revision 14 (June 1918) 300 200 'ii 1** :Ill ii

• f .... ... f .... I -100 ---""' ""'"' Drywell ""'"" ....... i.--"" .......... """ io-1oo ..... .. i.... "'"' ..... i-. ""'"' b-..... I--" "'"' Suppression Chamber [__.......... 102 103 104 NOTE: Time After Accident (sec) . A REVISED SUPPRESSION CHAMBER TEMPERATURE PROFILE EXISTS WHICH IS CONSISTENT WITH THE CONTAINMENT SPRAY SYSTEM OPERABILITY REQUIREMENTS FOR A MAXIMUM CONTAINMENT SPRAY RAW WATER TEMPERATURE Of s2*F. SEE FSAR SECTION XV*5.2 ) """"' """'" ..... loo looi. n 0 z -I > z 3:: m z -in m-1 3:: > .,, -mz ;:o 3: >m -tz c: -I ;:o 0 mm I (/) :e Cl -z -I :c ttJ o> c: V> -I-V> 8 ;:o m V> .,, ;:o > -<

c.,, .,, -cnC> >c :::XJ :::XJ m :xJ >< < -* I 0 0 ::s ):ii N 0 0 0 -l 0 (jJ m ?C !'-> 0 0 -.J LL <.!) uJ 0 DBR ANALYSIS SUPPRESSION POOL AND WETWELL AIRSPACE TEMPERATURE RESPONSE CONTAINMENT SPRAY DESIGN BASIS ASSUMPTION 300. 200. NMP-1 POOL HEATUP ANALYSJS RDL BREAK I SUPPRESION POOL rrEMP 2, AIRSPACE! TEMP I I I l I l 2 r I 2.. JJ er: 100. I-< er: UJ CL ::L uJ I-0. o. 1. 2. 3. 't. MINTZ 23600250 LOG TIME (SEC) 020305 0903.6 c .,, .,, -CJ) Ci) l> c :D :D >< < -* I O> 0 0 ::::s m N 0 0 ('\ -l 0 (jJ rn 7'\) N 8 ..... LL (Si LLJ 0 I DBR ANALYSIS SUPPRESSION POOL AND WETWELL AIRSPACE TEMPERATURE RESPONSE EOP OPERATION ASSUMPTIONS 300. 200. NMP-1 POOL HE.ATUP ANALYSl$ RDL BREAK l SUPPRESION POOL ITEMP Z. 1-JEn..IELL AIRSPACEITEMP I I I l 'l. 2. I/ I '2 LLJ c.r. 100. :::> f-< c.r. LLJ CL 2: LLJ f-0. o. l. 2. 3. "t. MJNTZ 2l002BOF LOG TIME (SEC) 02010'5 08'50.0 REACTOR BUILDING MODEL REACTOR BUILDING 43' MODEL FIGURE XV-61 UFIAR RevWon 14 (June 19M) ./ 4 .. vv 3 50 3 00 z: 0 i== < 0:: 2 LL. x UJ 50 00 50 00 50 l EXFILTRATION VS WIND SPEED -NORTHERLY WIND l I I I J I I v j I I I/ j I I I v / v 20 40 60 80 100 120 WIND SPEED (MPH) FIGURE XV-12 UflAll Revieion 14 (.June 19M) 0 -0.25 q, :t: *= -0.5 Q.. -0.7 0 5 "" REACTOR BUILDING DIFFERENTIAL PRESSURE r'-........ ...... \ \ ' II. \ \ \'{ORTHERLY WINO \ I\ \ \ \ \ \ \ \ \ \souTHERL y WIND I\ \ \ \ \ \ 20 40 60 80 100 120 WIND SPEED (MPH) FIGURE XV-13 UfSAR Rev181on 14 (June 1191) ( \ z: 0 j::: <( 20 0 18 0 16 0 14 0 12 0 .r " 10 ...J u... >< u.J 8 0 6 0 4 fU 2 0 l EXFILTRATION VS WIND SPEED -SOUTHERLY WIND ) I I I I I I I I I I I I ) / v 20 40 60 80 100 12 0 WIND SPEED (MPH) FfGUREXV-M UFIAR Aevilk>n 14 C.Jwte 1-) ( / Detail "A" See Figure XV-68 REACTOR BUILDING -ISOMETRIC Al I panels are caulked with H. H. Robertson RP 545 Compound Al I panel bolts are set in mastic FIGUAE XV-65 Uf*SAR Revision 14 (.June 1998) (. *-..._ \ / REACTOR BUI LOI NG -CORNER SECTIONS 7-5/8" 1 I ti 2, 4-5/8" GI RT LINE 9 ' -----+1-t COLUMN-SECTION A-A TYPICAL BETWEEN ELEV 398'-3! & 3951-3! I EXISTING 6 x 6 x 3/8" l SECTION B-B TYPICAL BETWEEN ELEV 395'-3! & ELEV 3841-0 (o) 7-5/8" GI RT LI NE T H SECTION C-C TYPICAL BETWEEN ELEV 384'-0 & ELEV 339'-lOt FIGURE XV-M UFSAR Revision 14 (June 1991) ( NUM CLOSER REACTOR BUILDING -ROOF SECTIONS Detail 11A11 r----12 GAUGE ALUMINUM SPLICE COVER 6" WIDE 12 GAUGE ALUMINUM COPING ALUMINUM BRACKET AND BACKER @2'-6 0/C STEEL CELL CAP CORRUGATED CLOSER 14 GAUGE METAL COATED INSULATION RETAINER ll"oilll+!l--H" FIBERGLASS INSULATION 18 GAUGE CORRUGATED PROTECTED METAL <BLACK) 9-11 /16" 18 GAUGE CORRUGATED PROTECTED METAL <BLACK) GAUGE METAL COATED Z BAR 0 I RIGID INSULATION (3/4" THICK) COMPOSITION BASE FLASHING 3k" CANT STRIP 3 PLIES 15# FINISH FELT 1 PLY 43# BASE FELT EDGE STRIP 2-6" TURBINE BUILDING NKX PANELS TYPICAL FOR TOP ELEV 398'-3t flQ.UflE xv Ml7 DRAIN UFW "8vision 14 (Ju.ne 1 HI) ., . ( REACTOR BUILDING -PANEL TO CONCRETE SECTIONS J 16 GAUGE CLOSURE (CONT. SEA U GIRT LINE SECTION G-G f'I") I co ........ Lr\ I r-TYPICAL BETWEEN ELEV 3391-10! & ELEV 336'-3! ALUMINUM FLASHING SEALANT ALUMINUM FLASHING *.

• 3/16" BOLTS @ 1 '-6c/c SET IN MASTIC EL.339'-10! ::.; . ..,.* \) . -.... -... ti*.. .-... Ii',_ SECTION H-H TYPICAL FOR ELEV 3391-10! AGURE XV-II UFIAR Revtston 14 (June 19M)

/ REACTOR BUI LOI NG -EXPANSION JOI NT SECTIONS 3'-3 1 '-9 1" 1 '-5' SECTION 0-0 TYPICAL BETWEEN ELEV 398'-3t & ELEV 395'-3t EXISTING 6 x 6 x 3/8" 1'-6-1/8 16 GAUGE EXPANSION CLOSURE (CONT. SEAL) ,..: -r------.. I I -+>=----H------1 I I 116 GAUGE CONT. ir I I 1 EXPANSION JOI NT SECTION E-E TYPICAL BETWEEN ELEV & ELEV 384'-0 K SECTION F-F rt"\ I GIRT LINE TYPICAL BETWEEN ELEV 384'-0 & ELEV 339'-lOt FIGURE XV-II UFSAR Revision 14 (June 11M} ( z 0 <C a:: 400 360 320 280 240 200 -Li... x w 160 12 0 8 0 4 0 0 REACTOR BUILDING EXFILTRATION -NORTHERLY WIND / !1 ) lj 71 I I I REVISED LEAV/ I /; VsUPPLEMENT 2 LEAKAGE /, I .V, I v I 20 40 60 100 WIND SPEED (MPH) FIGURE XV-70 UFIAR Revision 14 f.June 19", ( ( z 0 <( 200 180 160 140 120 g: 100 _I LL x lJ.J

• 80 60 40 20 l REACTOR BUI LOI NG EXFI LTRATION -SOUTHERLY WI ND /J II I 11 /;' I I/ /; I I 2 LEAKAGE REVISED lf I 20 40 60 80 100 WIND SPEED (MPH) FIGURE xv..;71 UPSAR Revision 14 CJune 1911)

Ct: LU < :;: (.fl * .:;. .. REACTOR BUILDING DIFFERENTIAL PRESSURE (.,.) z CL <l ...... Ct: LU < 3 WINO SPEED c (.fl ...... (.,.) z CL <l WINO SPEED FIGURE XV-72 UFSAR Rhtalon 14 (June 1191t .. *. 0 N I (/) Q) ,c 'U c Q) :J (/) (/) Q) CL Reactor Building Pressure vs. Time by Reactor Building Elevation 0.5 0.45 0.4 0.35 0.3 0.25 0.2 0.15 0.1 0.05 0 I -0.05 0.5 0 -0.1 -0.15 -0.2 -0.25 -----0.3 ,,__ _____ -0.35 -0.4 -0.45 -0.5 ---DP34C340 upper) ...._ -----DP35<340 lower) -DP36<318 upper) -------OP37C318 lower) -\... l .\ I I I I I I I I I I I I 1,5, 1 1.5 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 \ \ '. ' t\\ \ /67 minutes lower 340 elevation ' '-\\ '---/--5 hours upper 340 elevation \ ' \ "\ ---. \ ' --\ \ ---'-\ \ \ \ ----\\ -----------------------------------------------Time (hours) ------------FIGURE XV-73 UFSAR Rev1s1on 21 October-2009 Reactor Building Pressure vs. Time by Reactor Building Elevation CFocused on the Initial 2.5 hours) 0.5 0.45 0.4 0.35 I 0.3 0.25 0.2 ,,....., 0 0.15 N I 0.1 (f) Q) 0.05 ..c u 0 c I Q) -0.05 0.5 0 L -0.1 :::l (f) (f) -0.15 Q) L ()_ -0.2 -0.25 --0.3 -0.35 --------------------0.4 -0.45 -0.5 ---DP34C340 upper) -_,. __ ._ DP35C340 lower) -DP36<318 upper) -------DP37C318 lower) ..___ I I ':I /25.75 minutes upper 340 elevation , I I I ' 1' I I I I \I 0.5". 1 1.5 2 2.5 3 \\' " ' --" '-.... /67 minutes lower 340 elevation . ' . ' '-... "'-__/_ __ '-, ' ' ' -------, '-----.. -----------------------Time (hours) -------------FIGURE XV-74 UFSAR Rev1s1on 21 October 200C)

NMP Unit 1 UFSAR LIST OF EFFECTIVE FIGURES SECTION XVI Figure Revision Figure Revision Number Number Number Number Section XVI EF XVI-1 Rev. 25, October 2017 XVI-1 14 XVI-33 14 XVI-2 14 XVI-34 14 XVI-3 14 XVI-35 14 XVI-4 14 XVI-36 14 XVI-5 14 XVI-37 14 XVI-6 14 XVI-38 14 XVI-7 14 XVI-39 14 XVI-8 14 XVI-40 14 XVI-9 14 XVI-41 14 XVI-10 14 XVI-42 14 XVI-11 14 XVI-43 14 XVI-12 14 XVI-44 14 XVI-12a 14 XVI-45 14 XVI-12b 14 XVI-46 14 XVI-12c 16 XVI-47 14 XVI-12d 16 XVI-48 14 XVI-13 14 XVI-49 14 XVI-14 14 XVI-50 14 XVI-15 14 XVI-51 14 XVI-16 14 XVI-52 14 XVI-17 16 XVI-53 14 XVI-18 14 XVI-54 14 XVI-19 14 XVI-55 14 XVI-20 14 XVI-56 14 XVI-21 14 XVI-57 14 XVI-22 14 XVI-58 14 XVI-23 14 XVI-59 14 XVI-24 14 XVI-60 14 XVI-25 14 XVI-61 14 XVI-26 14 XVI-27 14 XVI-28 14 XVI-29 14 XVI-30 14 XVI-31 14 XVI-32 14 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section XVI XVI-i Rev. 25, October 2017 SECTION XVI SPECIAL TOPICAL REPORTS A. REACTOR VESSEL 1.0 Applicability of Formal Codes and Pertinent Certifications 2.0 Design Analysis 2.1 Code Approval Analysis 2.2 Steady-State Analysis 2.2.1 Basis for Determining Stresses 2.3 Pipe Reaction Calculations 2.4 Earthquake Loading Criteria and Analysis 2.4.1 Seismic Analysis for Core Shroud Repair Modification 2.5 Reactor Vessel Support Stress Design Criteria and Analysis 2.6 Strain Safety Margin for Reactor Vessels 2.6.1 Introduction 2.6.2 Strain Margin 2.6.3 Failure Probability 2.6.4 Results of Probability Analysis 2.6.5 Conclusions 2.7 Components Required for Safe Reactor Shutdown 2.7.1 Design Basis Load Combinations 2.7.2 Expected Stress and Deformation 2.7.2.1 Recirculation Line Break 2.7.2.2 Steam Line Break 2.7.2.3 Earthquake Loadings 2.7.3 Stresses and Deformations at Which the Component is Unable to Function and Margin of Safety 2.7.3.1 Recirculation Line Break 2.7.3.2 Steam Line Break 2.8 Safety Margins Against Ductile Fracture 3.0 Inspection and Test Report Summary 3.1 Materials 3.2 Fabrication and Inspection 4.0 Surveillance Provisions 4.1 Coupon Surveillance Program 4.2 Periodic Inspection 5.0 Core Shroud Repair Design NMP Unit 1 UFSAR Section Title Section XVI XVI-ii Rev. 25, October 2017 Description 5.1 Horizontal Weld Repair 5.2 Vertical Weld Repair B. PRESSURE SUPPRESSION CONTAINMENT 1.0 Applicability of Formal Codes and Pertinent Certifications 2.0 Design Analysis 2.1 Code Approval Calculations Under Rated Conditions 2.2 Ultimate Capability Under Accident Conditions 2.3 Capability to Withstand Internal Missiles and Jet Forces 2.4 Flooding Capabilities of the Containment 2.5 Drywell Air Gap 2.5.1 Tests and Inspections 2.6 Reactor Shield Wall 2.7 Compatibility of Dynamic Deformations Occurring in the Drywell, Torus, and Connecting Vent Pipes 2.8 Containment Penetrations 2.8.1 Classification of Penetrations 2.8.2 Design Bases 2.8.3 Method of Stress Analysis 2.8.4 Leak Test Capability 2.8.5 Fatigue Design 2.8.6 Material Specification 2.8.7 Applicable Codes 2.8.8 Jet and Reaction Loads 2.9 Drywell Shear Resistance Capability and Support Skirt Junction Stresses 3.0 Inspection and Test Report Summary 3.1 Fabrication and Inspection 3.2 Tests Conducted 3.3 Discussion of Results 3.3.1 Results 3.3.2 Effect of Various Transients 3.3.2.1 Ambient Temperature and Solar Heating of Shell 3.3.2.2 Thermal Lag Through Reference Chamber Wall NMP Unit 1 UFSAR Section Title Section XVI XVI-iii Rev. 25, October 2017 3.3.2.3 Condensation in Reference Chamber 3.3.2.4 Volume Changes Due to Thermal Transients 3.3.2.5 Overpressure Test--Plate Stresses C. ENGINEERED SAFEGUARDS 1.0 Seismic Analysis and Stress Report 1.1 Introduction 1.2 Mathematical Model 1.3 Method of Analysis 1.3.1 Flexibility or Influence Coefficient Matrix 1.3.2 Normal Mode Frequencies and Mode Shapes 1.3.3 The Seismic Spectrum Values 1.3.4 Dynamic Modal Loads 1.3.5 Modal Response Quantities 1.3.6 The Combined Response Quantities 1.3.7 Basic Criteria for Analysis 1.4 Discussion of Results 2.0 Containment Spray System 2.1 Design Adequacy at Rated Conditions 2.1.1 General 2.1.2 Condensation and Heat Removal Mechanisms 2.1.3 Mechanical Design 2.1.4 Loss-of-Coolant Accident 2.2 Summary of Test Results 2.2.1 Spray Tests Conducted 3.0 Core Spray and Containment Spray Suction Strainers D. DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 1.0 Classification and Seismic Criteria 1.1 Design Techniques 1.1.1 Structures 1.1.2 Systems and Components 1.2 Pipe Supports 1.3 Seismic Exposure Assumptions 2.0 Plant Design for Protection Against Postulated Piping Failures in High-Energy Lines 2.1 Inside Primary Containment NMP Unit 1 UFSAR Section Title Section XVI XVI-iv Rev. 25, October 2017 2.1.1 Containment Integrity Analysis 2.1.1.1 Fluid Forces 2.1.1.2 Impact Velocities and Effects 2.1.2 Systems Affected by Line Break 2.1.3 Engineered Safeguards Protection 2.2 Outside Primary Containment 3.0 Building Separation Analysis 4.0 Tornado Protection 5.0 Thermally-Induced Overpressurization of Isolated Piping E. EXHIBITS F. CONTAINMENT DESIGN REVIEW G. USAGE OF CODES/STANDARDS FOR STRUCTURAL STEEL AND CONCRETE H. REFERENCES NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section XVI XVI-v Rev. 25, October 2017 XVI-1 CODE CALCULATION SUMMARY XVI-2 STEADY-STATE - (100% FULL POWER NORMAL OPERATION) PERTINENT STRESSES OR STRESS INTENSITIES XVI-3 LIST OF REACTIONS FOR REACTOR VESSEL NOZZLES XVI-4 EFFECT OF VALUE OF INITIAL FAILURE PROBABILITY XVI-5 SINGLE TRANSIENT EVENT FOR REACTOR PRESSURE VESSEL XVI-6 POSTULATED EVENTS XVI-7 MAXIMUM STRAINS FROM POSTULATED EVENTS XVI-8 CORE STRUCTURE ANALYSIS RECIRCULATION LINE BREAK XVI-9 CORE STRUCTURE ANALYSIS STEAM LINE BREAK XVI-9a CORE SHROUD REPAIR DESIGN SUPPORTING DOCUMENTATION XVI-10 DRYWELL JET AND MISSILE HAZARD ANALYSIS DATA XVI-11 DRYWELL JET AND MISSILE HAZARD ANALYSIS RESULTS XVI-12 STRESS DUE TO DRYWELL FLOODING XVI-13 ALLOWABLE WELD SHEAR STRESS XVI-14 LEAK RATE TEST RESULTS XVI-15 OVERPRESSURE TEST--PLATE STRESSES XVI-16 STRESS SUMMARY XVI-17 HEAT TRANSFER COEFFICIENTS AS A FUNCTION OF DROP DIAMETER XVI-18 HEAT TRANSFER COEFFICIENT AS A FUNCTION OF PRESSURE NMP Unit 1 UFSAR Table Number Title Section XVI XVI-vi Rev. 25, October 2017 XVI-19 RELATIONSHIP BETWEEN PARTICLE SIZE AND TYPE OF SPRAY PATTERN XVI-20 ALLOWABLE STRESSES FOR FLOOR SLABS, BEAMS, COLUMNS, WALLS, FOUNDATIONS, ETC. XVI-21 ALLOWABLE STRESSES FOR STRUCTURAL STEEL XVI-22 ALLOWABLE STRESSES - REACTOR VESSEL CONCRETE PEDESTAL XVI-23 DRYWELL - ANALYZED DESIGN LOAD COMBINATIONS XVI-24 SUPPRESSION CHAMBER - ANALYZED DESIGN LOAD COMBINATIONS XVI-25 ACI CODE 505 ALLOWABLE STRESSES AND ACTUAL STRESSES FOR CONCRETE VENTILATION STACK XVI-26 ALLOWABLE STRESSES FOR CONCRETE SLABS, WALLS, BEAMS, STRUCTURAL STEEL, AND CONCRETE BLOCK WALLS XVI-27 SYSTEM LOAD COMBINATIONS XVI-28 HIGH-ENERGY SYSTEMS - INSIDE CONTAINMENT XVI-29 HIGH-ENERGY SYSTEMS - OUTSIDE CONTAINMENT XVI-30 SYSTEMS WHICH MAY BE AFFECTED BY PIPE WHIP XVI-31 CAPABILITY TO RESIST WIND PRESSURE AND WIND VELOCITY NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section XVI XVI-vii Rev. 25, October 2017 XVI-1 SEISMIC ANALYSIS OF REACTOR VESSEL GEOMETRIC AND LUMPED MASS REPRESENTATION XVI-2 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. MOMENT XVI-3 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. SHEAR XVI-4 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. DEFLECTION XVI-5 REACTOR SUPPORT DYNAMIC ANALYSIS - ELEVATION VS. ACCELERATION XVI-6 thru FIGURES DELETED XVI-8 XVI-9 REACTOR VESSEL SUPPORT STRUCTURE STRESS SUMMARY XVI-10 THERMAL ANALYSIS XVI-11 FAILURE PROBABILITY DENSITY FUNCTION XVI-12 ADDITION STRAINS PAST 4% REQUIRED TO EXCEED DEFINED SAFETY MARGIN XVI-12a SHROUD WELDS XVI-12b CORE SHROUD STABILIZERS XVI-12c CORE SHROUD WELDS XVI-12d V9/V10 VERTICAL WELD CLAMP ASSEMBLY XVI-13 LOSS-OF-COOLANT ACCIDENT - CONTAINMENT PRESSURE NO CORE OR CONTAINMENT SPRAYS XVI-14 FIGURE DELETED XVI-15 DRYWELL TO CONCRETE AIR GAP XVI-16 TYPICAL PENETRATIONS NMP Unit 1 UFSAR Figure Number Title Section XVI XVI-viii Rev. 25, October 2017 XVI-17 REACTOR SHIELD WALL CONSTRUCTION DETAILS XVI-18 VENT PIPE AND SUPPRESSION CHAMBER XVI-19 PRIMARY CONTAINMENT SUPPORT AND ANCHORAGE XVI-20 SEAL DETAILS - DRYWELL SHELL STEEL AND ADJACENT CONCRETE XVI-21 DRYWELL SLIDING - ACCELERATION, SHEAR, AND MOMENT XVI-22 SHEAR RESISTANCE CAPABILITY - INSIDE DRYWELL XVI-23 SHEAR RESISTANCE CAPABILITY - OUTSIDE DRYWELL XVI-24 DRYWELL - SUPPORT SKIRT JUNCTION STRESSES XVI-25 POINT LOCATION FOR CONTAINMENT SPRAY SYSTEM PIPING HEAT EXCHANGER TO DRYWELL XVI-26 COMPARISON OF STATIC AND DYNAMIC STRESSES (PSI) SEISMIC CONDITIONS - CONTAINMENT SPRAY SYSTEM HEAT EXCHANGER TO DRYWELL XVI-27 CONDUCTION IN A DROPLET XVI-28 LOSS OF COOLANT ACCIDENT - CONTAINMENT PRESSURE XVI-29 LOSS OF COOLANT ACCIDENT - CONTAINMENT PRESSURE XVI-30 NOZZLE SPRAY TEST - PRESSURE DROP OF 80 PSIG XVI-31 NOZZLE SPRAY TEST - PRESSURE DROP OF 80 PSIG XVI-32 NOZZLE SPRAY TEST - PRESSURE DROP OF 30 PSIG XVI-33 NOZZLE SPRAY TEST - PRESSURE DROP OF 30 PSIG XVI-34 SEISMIC ANALYSIS - REACTOR BUILDING XVI-35 DYNAMIC ANALYSIS - DRYWELL NMP Unit 1 UFSAR Figure Number Title Section XVI XVI-ix Rev. 25, October 2017 XVI-36 REACTOR SUPPORT STRUCTURE - SEISMIC XVI-37 SEISMIC ANALYSIS - WASTE BUILDING XVI-38 SEISMIC ANALYSIS - SCREENHOUSE XVI-39 SEISMIC ANALYSIS - TURBINE BUILDING (NORTH OF ROW C) XVI-40 SEISMIC ANALYSIS - TURBINE BUILDING (SOUTH OF ROW C) XVI-41 SEISMIC ANALYSIS - CONCRETE VENTILATION STACK XVI-42 REACTOR BUILDING MATHEMATICAL MODEL (NORTH-SOUTH) XVI-43 REACTOR SUPPORT STRUCTURE - SEISMIC XVI-44 REACTOR SUPPORT STRUCTURE - REACTOR BUILDING XVI-45 REACTOR SUPPORT STRUCTURE - REACTOR BUILDING AND SEISMIC XVI-46 PLAN OF BUILDING XVI-47 WALL SECTION 1 XVI-48 WALL SECTION 1 - DETAIL "A" XVI-49 WALL SECTION 1 - DETAIL "B" XVI-50 WALL SECTION 1 - DETAIL "C" XVI-51 WALL SECTION 1 - DETAIL "D" XVI-52 WALL SECTION 1 - DETAIL "E" XVI-53 WALL SECTION 2 XVI-54 WALL SECTION 3 XVI-55 WALL SECTION 3A - DETAILS NMP Unit 1 UFSAR Figure Number Title Section XVI XVI-x Rev. 25, October 2017 XVI-56 WALL SECTION 4 XVI-57 WALL SECTION 4 - DETAIL 1 XVI-58 WALL SECTION 4 - DETAIL 2 XVI-59 WALL SECTION 5 XVI-60 WALL SECTION 6 XVI-61 WALL SECTION 7 NMP Unit 1 UFSAR Section XVI XVI-1 Rev. 25, October 2017 SECTION XVI SPECIAL TOPICAL REPORTS A. REACTOR VESSEL 1.0 Applicability of Formal Codes and Pertinent Certifications The Nine Mile Point Nuclear Station - Unit 1 (Unit 1) reactor vessel was fabricated, inspected, and tested in accordance with the ASME Boiler and Pressure Vessel Code, Section I, Power Boilers, 1962 Edition and Addenda plus the Nuclear Code Cases applicable on December 11, 1963. This was the date of contract between the buyer, General Electric Company Atomic Power Equipment Department (GE APED), San Jose, California, and the seller, Combustion Engineering, Inc., Chattanooga Division, Chattanooga, Tennessee. The vessel purchase specification further directed the seller to use Section VIII, Unfired Pressure Vessels, where Section I, Power Boilers, did not cover specific details. The above Boiler and Pressure Vessel Sections and Cases were used: 1. To establish minimum shell, head, flange and nozzle material thickness. 2. As a base for establishing the inspections and tests required by the State of New York and any local governing bodies. 3. As a base for formulation of further design rules, inspections and tests not covered by the Sections and Cases. The vessel purchase specification, in addition to the Code sizing calculations, required the seller to submit a detailed stress analysis covering both steady-state and transient analysis with respect to material fatigue. The summary of results of the detailed stress analysis is contained in Section A.2.2. Inspections and tests, in addition to those required by the Code, were required by the specifications which included the following pertinent inspections and/or tests. NMP Unit 1 UFSAR Section XVI XVI-2 Rev. 25, October 2017 1. Establish specific maximum nil ductility transition temperatures (NDTT) for vessel shell material opposite the core region and elsewhere (40°F NDTT). 2. A fabrication test program on vessel shell material which included testing of large size tensile specimens (80 percent of the vessel wall thickness in diameter), both plain and welded samples. 3. Provisions were made for determining the effects of nuclear radiation upon the reactor vessel structural materials by supplying specimens of the vessel material to be exposed to the core irradiation at the vessel wall inside of the vessel. Section III, Nuclear Vessels Code, was not used as this Section was not approved for general use until after the contract date for purchase of the reactor vessel. The reactor vessel was ordered in December 1963 and Section III was not released until 1964. Pertinent certifications are contained in a report by Combustion Engineering, Inc., Inspection Report for the 213 Inch I. D. Reactor Vessel and Components, CE Serial Nos. 64101 and 64201, CE Contract 164. The following are included in the report: (1) Materials Certification (2) Hydrostatic Certification (3) Radiographic Certification (4) Liquid Penetrant Certification (5) Ultrasonic Certification (6) Magnetic Particle Certification (7) Cleaning Certification (8) Welding Certification (9) Weld Repair Certification (10) Furnace Logs (11) List of Piece Numbers, Code Numbers, and Heat Numbers (12) Mill Test Reports and Certifications (13) As-Built Drawings (14) Material Identification Chart (15) X-Ray Location Drawings (16) Name Plate Photographs (17) ASME Data Report Form P-3 (18) Weld Rod and Wire Certification (19) Carbon Content of Clad Certification NMP Unit 1 UFSAR Section XVI XVI-3 Rev. 25, October 2017 (20) Cobalt Content of Clad Certification (21) Stress Relieving Certification (22) Welders Qualification Certification (23) Parkerizing Certification (24) Quench and Temper Certification There were no deviations to the formal Codes throughout the design, fabrication, inspection and testing of the reactor vessel. 2.0 Design Analysis 2.1 Code Approval Analysis Analyses were performed to demonstrate that the reactor vessel met the requirements of the specified Codes. Table XVI-1 compares Code required thicknesses and reinforcing areas with actual values. 2.2 Steady-State Analysis The stresses in the reactor vessel under maximum steady-state loads, including dead, live, and seismic loads, do not exceed the Code allowable stresses as shown in Table XVI-2. Stresses were determined in accordance with the GE vessel purchase specification. 2.2.1 Basis for Determining Stresses The determination of stresses under this specification is based on the maximum shear theory. Stresses are expressed in terms of the "stress intensity" which is defined as twice the maximum shear stress, or as the largest algebraic difference between any two of the three principal stresses at a given point. Terms relating to stress determination which are used in this specification are defined as follows: Membrane Stress The component of normal or direct stress which is uniformly distributed and equal to the average value of stress across the section under consideration away from structural discontinuities. Local Membrane Stress The component of normal or direct stress which is uniformly distributed and equal to the average value of stress across the section under consideration at a structural discontinuity. NMP Unit 1 UFSAR Section XVI XVI-4 Rev. 25, October 2017 Bending Stress The component of normal or direct stress which varies with the distance from the centroid of the section under consideration. Primary Stress A direct stress or shear stress developed by the imposed loading which is necessary to satisfy the simple laws of equilibrium of external and internal forces and moments. The basic characteristic of a primary stress is that it is not self-limiting. In the absence of strain hardening, a primary stress which considerably exceeds the yield strength will result in failure, or at least gross distortion. A thermal stress is never classified as a primary stress. Examples of primary stresses are: 1. Membrane stress in a circular, cylindrical or spherical shell due to internal pressure. 2. Bending stress in the central portion of a flat head due to pressure. Secondary Stress A direct stress or shear stress developed by the constraint of adjacent parts or by self-constraint of a structure. The basic characteristic of a secondary stress is that it is self-limiting. Local yielding and minor distortions can satisfy the conditions which cause the stress to occur, and failure from one application of the stress is not to be expected. Examples of secondary stress are: 1. Thermal stress 2. Bending stress at a gross structural discontinuity Gross Structural Discontinuity A source of strain or stress intensification which affects a relatively large portion of a structure and has a significant effect on the overall stress or strain pattern in the structure as a whole. Examples are head-to-shell and flange-to-shell junctions and junctions between shells of different diameters or thicknesses. Local Structural Discontinuity A source of stress or strain intensification which affects a relatively small volume of material and does not have a significant effect on the overall stress or strain pattern nor on the structure as a whole. Examples are opening and nozzle connections, small fillet radii, small attachments, and incomplete weld penetrations. NMP Unit 1 UFSAR Section XVI XVI-5 Rev. 25, October 2017 2.3 Pipe Reaction Calculations Reactions at the reactor vessel nozzles are due mainly to thermal stress conditions. The stress analyses were calculated by computer with a program based on vector and matrix algebra. The results are shown in Table XVI-3. 2.4 Earthquake Loading Criteria and Analysis Loads imposed by the design earthquake were taken into account in the design of the reactor vessel.(1-3) Figure XVI-1 shows the arrangement and configuration of the reactor pressure vessel (RPV), supporting concrete pedestal, and shield wall. The maximum dynamic moment envelope for the RPV supporting pedestal is shown on Figure XVI-2. The moments shown are in kip-feet units. Values are the moments that result from the seismic response of the RPV alone plus the effect of reactor building movement. Figure XVI-3 shows the maximum dynamic shear envelope on the RPV concrete pedestal and includes the effect of building movement. The maximum dynamic displacement envelope is shown on Figure XVI-4 and includes the effect of building movement. The maximum dynamic acceleration envelope is shown on Figure XVI-5. A jet load analysis was also performed with the results illustrated in Figures XVI-2, XVI-3, and XVI-4. 2.4.1 Seismic Analysis for Core Shroud Repair Modification A dynamic seismic analysis(4) was performed for the reactor core shroud repair modification. The mathematical, beam element structural model used for the seismic analysis includes the reactor building, shield wall/pedestal, RPV, reactor internals, and the repair modification hardware, all coupled. The model was analyzed using the SAP4G07 computer program. The current licensing basis design basis earthquake (DBE) was used in the analysis. A synthetic time-history was generated based on the horizontal DBE spectra in accordance with the guidelines contained in NUREG-0800. DBE results were combined NMP Unit 1 UFSAR Section XVI XVI-6 Rev. 25, October 2017 with upset as well as emergency and faulted conditions. To add conservatism to the shroud repair design, the DBE loads were combined with the loss-of-coolant accident (LOCA) loads, although the original plant licensing basis did not require this load combination. No detailed RPV and internals dynamic analysis documentation existed for Unit 1 against which the results of the new detailed dynamic model could be benchmarked. The new coupled model was thus generated utilizing the available licensing basis data and analyzed and verified using the methodologies employed in modern plant designs. The licensing basis condition was, however, simulated by additionally analyzing the model without the shroud stabilizers and without any cracks, to form a new benchmark run. The resultant component loads based on the new shroud repair seismic analysis compared favorably with the component loads in the benchmark run. 2.5 Reactor Vessel Support Stress Design Criteria and Analysis The reactor vessel supports were analyzed taking into account thermal stresses in addition to all other loading conditions as tabulated on Figure XVI-9. The variation in temperature of these structures, as determined by the thermal analysis, is shown on Figure XVI-10. The condition causing the most rapid change of temperature in the reactor vessel base occurs during a normal startup, cooldown, or scram condition. The maximum temperature differential between the reactor vessel base and the ring girder, and between the ring girder and support concrete, is 15°F. This is a result of the formation of a nonventilated air pocket in the upper section of the concrete support pedestal above the ventilation opening at el 252-2. The lower section of the concrete support pedestal below the ventilation openings at el 252-2 is cooled by the six drywell cooling fans at the bottom of the drywell. Ventilation openings in the reactor support pedestal at an upper and lower level will maintain a 5°F temperature differential between the inside and outside of this structure. Thermal stresses resulting from the temperature differentials are shown on Figure XVI-9. The radial growth based on the maximum temperature differential is 0.01025 in, resulting in a reactor vessel base-bolt shear stress of 2430 psi (Figure XVI-9), or a required increased friction factor of 0.0158. NMP Unit 1 UFSAR Section XVI XVI-7 Rev. 25, October 2017 Since for the greater part the ring girder is embedded in grout and adjacent to large concrete masses, its temperature will remain close to that of the surrounding concrete. For a maximum temperature differential of 15°F for the above conditions, the resulting radial thermal growth of the ring girder produces a concrete friction or shear stress of 45 psi (see point A1 on Figure XVI-9). Temperature distribution in the reactor vessel support foot shows the temperature to be 125°F in the foot and 465°F in the attachment of the vessel support skirt to the vessel (Figure XVI-10). Stress levels in the skirt remain within allowable limits for all loading conditions. The condition causing the most rapid change of ambient temperature is a LOCA (recirculation line break). The resulting jet thrust would last approximately 15 sec. At this time there would be essentially no temperature change in any of the reactor vessel support structure components. Subsequent reactor vessel loadings would not have jet thrust or operating loadings combined with the dead and earthquake loadings. Reactor vessel base-bolt stresses for the dead load and earthquake condition are 38,000 psi tension and 8,900 psi shear with no shear stress due to radial thermal growth. Using the combined shear and tension stress equation for high-strength bolts in the AISC Handbook, Sixth Edition (1963)*, the total allowable bolt shear stress of 13,700 psi will allow a thermal radial differential of 30°F between the reactor vessel base and the ring girder. Raising allowable stress levels to 0.9 of functional failure will allow total reactor vessel bolt shear stresses of 29,900 psi, which results in an allowable radial thermal differential of 185°F. The horizontal shear load for the earthquake condition is transferred from the ring girder to the support concrete in bearing on the shear key below the shield wall. The bearing stress shown on Figure XVI-9, point A2, is reduced from 725 psi to 670 psi for this loading condition. Total horizontal shear resistance at this level utilizing the shear key, friction or shear surface, and ring girder anchor bolts allows an earthquake and radial thermal growth loading caused by a temperature differential of 90°F with no stress increase above normal Code values. If stress levels of 0.9 of functional failure are used, this allowable temperature differential is 178°F. Also see Section XVI, Subsection G. NMP Unit 1 UFSAR Section XVI XVI-8 Rev. 25, October 2017 The concrete reactor vessel support structure can accept a 45°F temperature differential combined with operating, earthquake, and jet thrust loadings before normal allowable stress values, as stated in ACI and AISC handbooks, are reached. If stress levels of 0.9 of functional failure are used, this allowable temperature differential is 110°F. For the accident condition wherein all drywell coolers are lost, ambient temperature changes will result in thermal radial growth of the reactor vessel support structure components. The radial growth caused by an assumed differential temperature rise of 110°F in these components would produce stresses below 0.9 of functional failure when combined with the earthquake and jet thrust loadings. It is not conceivable that ambient temperature changes could produce a temperature differential of this magnitude in these components. The original thermal analysis inside the drywell shows a maximum operating temperature in the drywell of 110°F at the bottom and 135°F at the top. 2.6 Strain Safety Margin for Reactor Vessels 2.6.1 Introduction Thermal stresses occur in a vessel when two segments or areas are at different temperatures. The thermal stresses and strains on the reactor vessel or reactor system which can result from system operation are limited in order to prevent fatigue or distortion of the vessel. The thermal strains are controlled by limitations on allowed heatup and cooldown rates, and limitations on cold inlet water temperatures (and thermal sleeves). The expected thermal strains on the vessel were analyzed and included in the fatigue analysis. It was concluded from the fatigue analysis that the vessel could withstand thermal strains beyond those anticipated during normal operation. In fact, conditions well beyond the design limits would be required to produce vessel failure. Some types of stresses or strains are more critical than others. For example, a system overpressure (beyond the design or test pressure), which is nonrelieving primary stress, would be considered a more serious event than a severe thermal transient whose calculated maximum local strain range exceeded the Code design limit. In either case an elastic-plastic analysis is necessary to establish the level of local strain. In addition, NMP Unit 1 UFSAR Section XVI XVI-9 Rev. 25, October 2017 the unexpected event may have a different distribution of strains than those due to the design events, so that total material damage accumulation at any one point would affect the concern about a single overdesign event. Therefore, a method for evaluating the vessel damage due to thermal transients and a safety margin on vessel strain were established. For RPVs designed and built prior to the adoption of the ASME Boiler and Pressure Vessel Code Section III, Nuclear Vessels, GE developed a method for making a fatigue analysis which provided assurance that vessels installed in GE design nuclear power plants would safely withstand all anticipated operating and transient conditions, both normal and emergency. This method was based upon the method of analysis developed for Naval reactors and upon industry's experience using it. This method was defined by and was a specific requirement of the design specification for the Nine Mile Point reactor vessel, in addition to the requirements for compliance with Section I, Section VIII, and applicable Nuclear Code Cases. Section III now includes the requirements of the Code cases and the method of analysis with only minor changes. However, the allowable stresses under Section III are greater in the areas of concern than were permitted by Sections I and VIII and the fatigue curves used in the design specification. Therefore, the use of Section I and Section VIII, plus the Nuclear Code Cases, plus the GE Specification-defined analysis, results in a completed vessel for the Nine Mile Point Nuclear Station which has strain safety margins that are, in general, equivalent to those which would have resulted from using Section III. A quantitative safety margin concept can be established by considering the failure risk. This failure risk can be compared to that inherent in the method required by the design specification and the ASME Section III Pressure Vessel Code. It is expected that such risk due to thermal transients would affect only the fatigue life of the vessel; that is, it would only be expected to increase the risk of cracking or leaking, but not have any significant effect on the gross strength of the vessel to contain pressure. 2.6.2 Strain Margin The design criteria included in the reactor vessel specifications for Nine Mile Point which, as stated previously, are in essential agreement with the criteria subsequently incorporated in the ASME Code, Section III, provide assurance that a vessel designed, built and operated within its design NMP Unit 1 UFSAR Section XVI XVI-10 Rev. 25, October 2017 limits will have an extremely low probability of failure due to any known failure mechanism. The design specification methods and design analysis include rare but expected events which may occur during any part of the vessel lifetime. For a carbon steel vessel, even the ASME Section III criteria allow for a local strain range during a transient (including strain concentration effects) of about 4.0 percent to occur as many as ten times during the vessel lifetime. Somewhat greater strains than this, or more frequent cycles of this strain, would not be expected to fail the pressure vessel because a "safety factor" was included in the design limit. An extrapolation of the low-cycle fatigue design criteria curve, which was a part of the design specification for the Nine Mile Point vessel for carbon steel, would result in an allowable strain which would be larger than the design specification even for a single cycle, because of the difficulty in obtaining materials data in this region and the difficulty in making accurate plastic strain predictions in this high-strain region. However, such a strain is appropriate to evaluate the desired safety margin. 2.6.3 Failure Probability Failure can be treated as a probability phenomenon with the peak strain range considered as a random variable. Many other important factors, such as fabrication and inspection methods and material toughness, play a major role in establishing pressure vessel reliability; however, the peak strain is the most universal single parameter which can be used in establishing a safety margin. Strain is treated as a random variable because of the uncertainty surrounding the value of strain at which failure would occur. At levels as low as 4.0 percent it is almost certain that failure would not occur, and at a level of 35 to 40 percent strain failure would occur. The 35 to 40 percent region was established from the ASME Section III criteria by extrapolating the lower limit of actual low-cycle fatigue data from ten cycles to one cycle (Figure 10 of "Criteria of Section III of ASME Boiler and Pressure Vessel Code for Nuclear Vessels"). A probability density function for failure can be constructed in the form shown on Figure XVI-11. This curve implies that a small but finite risk of failure occurs at any strain above 4.0 percent, but that a much larger strain would be required to make the probability of failure close to unity. The 4.0 percent value is arbitrary, but is based on much successful experience in preventing the initiation NMP Unit 1 UFSAR Section XVI XVI-11 Rev. 25, October 2017 of cracking. At what strain does the probability of failure change from insignificant (0) to significant (>0)? For this evaluation the important portion of the density function is the region near 4.0 percent. It is essential to fit a density function to this problem which does not go to zero even at 4.0 percent. There are many possible theoretical density functions which can be used to fit this problem. With a linear change of variable of the form: Where: x = Transformed random variable = Strain at which failure is known to occur o = Original strain to failure random variable the density functions of the chi-square or gamma variety are reasonable.* In terms of the changed strain variable, the failure probability can be expressed as: Where: n = Chi-square parameter usually referred to as "degree-of-freedom," chosen to fit the problem = Gamma function This distribution has a finite probability even for a negative strain o, but this zone is small enough to exert negligible effect on the result being established. The procedure to establish the difference between nearly zero failure probability at 4.0 percent strain and a nonzero failure probability, which This distribution, used in this way, admits below average strength behavior, but not unrealistically large values of strength. There is no "unique distribution" which fits this problem, and the "Baysian" conception of probability has been adopted here. NMP Unit 1 UFSAR Section XVI XVI-12 Rev. 25, October 2017 is of safety concern at some higher strain, will be to: 1) arbitrarily assign a small but finite probability to the 4.0 percent level, and 2) determine the strain at which the arbitrarily small number increases an order of magnitude. For example, if 10-6 is assigned as the 4.0 percent strain risk, the strain at which the risk becomes 10-5 is defined as the safety margin. As long as the initial number is sufficiently small, 10 times that number is still a very small failure probability, and yet a numerical procedure is available to evaluate the risk for any strain which could be hypothesized. The choice of 10-6 as the starting point is a reasonable choice for the starting point, based on Borel's scale of zero probability for "human" probability scales**, such as in these design procedures. 2.6.4 Results of Probability Analysis For convenience in the probability tables, an of 38 was selected. For the 4.0 percent strain, a value of n = 4 produces Pf = 0.9 x 10-6; the value of o at which Pf = 0.9 x 10-5 is found to be 8.0 percent. To demonstrate the sensitivity of the result to the assumption of the initial arbitrary Pf of 10-6, the procedure is repeated for two more initial points, one larger and one smaller. The strain at which the initial risk increases by an order of magnitude is calculated based on an of 38 (see Table XVI-4). The results are not very sensitive to the initial Pf chosen and, in fact, the 8.0 percent result is a minimum over this range. Not too much significance can be attached to this minimum because this family of distributions has a variable mean. Since we have very little knowledge of what the mean is, there is little way to judge more quantitatively the distributions chosen. 2.6.5 Conclusions These additional strains (past 4.0 percent) required to exceed the defined safety margin are plotted together with the result of extrapolating the actual fatigue data from ASME Section III from 10 cycles to 1 cycle (Figure XVI-12). Therefore, the vessel could withstand an additional 4.0 percent strain beyond even the ASME Section III fatigue limit for the single cycle event with still very small failure probability. In summary, the single cycle overdesign transient safety margin is as shown in Tables XVI-5, XVI-6 and XVI-7. Measurement: Definitions and Theories edited by Churchman and Ratsosh. NMP Unit 1 UFSAR Section XVI XVI-13 Rev. 25, October 2017 The equivalent effect strain range is implied here using the shear theory of combined strains and using the definition of strain range in the design specification and later incorporated in the ASME Section III Code. The strains from potential operating events (Table XVI-6) are compared with the safety margins in Table XVI-7 of elastic extrapolated strains. It should be pointed out that these values are very approximate and that more precise evaluations generally require an elastic plastic-analysis for each zone and for each postulated event. 2.7 Components Required for Safe Reactor Shutdown The following components must maintain their functional capabilities to ensure safe reactor shutdown: Lower shroud and core support Upper shroud Core support plate Guide tubes Fuel channel Top guide Core shroud stabilizers* Core shroud vertical weld repair clamps* Functional capability means limiting deformation and deflection so that components maintain their cage-like configuration to the extent that design control rod drive (CRD) scram capability and design core spray functions are not affected. 2.7.1 Design Basis Load Combinations For each of the above components the design basis load combination was as follows: For the normal operating transients and steady-state condition, the above components and the internal components meet all the primary and secondary stress requirements of Section III, ASME The core shroud stabilizers and vertical weld repair clamps were not a part of the original reactor internal components required to ensure safe reactor shutdown. The shroud stabilizers were installed in the reactor in 1995 and the vertical weld repair clamps were installed in 1999. Sections 2.7.1 through 4.2 have not been revised to account for the new stabilizers. Instead, Sections IV-A.7.1.10, IV-B.7.1.9 and XVI-A.5.0 were added to the UFSAR to provide design details of the core shroud repair and/or the vertical weld repair clamps. The analyses listed in Table XVI-9a provide the calculated displacements of the various core structure components under earthquake and recirculation/main steam line break loadings. NMP Unit 1 UFSAR Section XVI XVI-14 Rev. 25, October 2017 Code for Class "A" Vessels. To ensure that these stress requirements were met, the design earthquake loading was combined with the loads induced during normal operating transients and steady-state conditions. For the accident loads, deflections and deformation were limited under an accident loading to maintain the functional capability of all structural components. 2.7.2 Expected Stress and Deformation 2.7.2.1 Recirculation Line Break The expected stress and/or collapse loads (if buckling is a mode of failure) are given in Table XVI-8. The top guide is not included in the tabulation because it is an open "egg crate" that literally supports the fuel and is not a pressure barrier; therefore, the maximum accident loadings produce negligible stresses on the structure. Deformations are not listed in the table but are negligible (well within elastic limits), except on the core support plate and fuel channel. Total deflection at the center of the core support plate is about 1 3/4 in, with a permanent deformation of about 1 1/2 in. The deformation does not in any way affect the design scram capability of the CRD. Stress is not tabulated for the core support plate because the surface stress is above yield. Note that this does not mean that uncontrolled deformation occurs. Only the top and bottom surface stresses are above the yield; the stresses in the interior of the beam are still within the elastic limit. During a recirculation line break (immediate reactor scram), the fuel channels will possibly buckle inward at the bottoms of the assemblies. The pressure reported in Table XVI-8 occurs at the bottom of a given channel, whereas no differential pressure is applied at the channel top. The inward motion of the channels is limited to about 0.140 in by the fuel bundle assembly, which is, in turn, substantially supported along its length by the inlet casting and the fuel spacers. Because of the inward movement, the clearance between the control rod and the fuel channel will be increased; hence, the deformation will not hamper insertion of the control rod. NMP Unit 1 UFSAR Section XVI XVI-15 Rev. 25, October 2017 2.7.2.2 Steam Line Break The applied differential pressure and the resultant stresses on the various internal components due to a steam line break are listed in Table XVI-9. The differential pressures listed in Table XVI-9 were derived assuming the main steam line break (MSLB) occurs at initial conditions of 100 percent power and 100 percent core flow. If the break occurs at certain low power, high flow initial conditions, then the upper shroud differential pressure will be higher than the value listed in Table XVI-9. This low power, high flow MSLB is typically referred to as the faulted interlock point and is defined on the power flow maps as the reactor internals protection (RIP) region. Avoidance of this low power, high flow region is an initial assumption for the MSLB accident analysis with respect to the evaluation of structural integrity of the vessel internals. The RIP region is defined by the following boundaries on the power flow maps; at 65 percent rated core flow a line from 0 percent power up to 20 percent rate power, a line connecting 65 percent core flow at 20 percent power, and 100 percent core flow at 50 percent power. Administrative controls to restrict operation in the RIP region provide assurance that the upper shroud differential pressure listed in Table XVI-9 will not be exceeded in the event of a MSLB. Deformation of the structural components listed in Table XVI-9 are negligible except for the fuel channel. Total outward deformation due to a 30-psi internal pressure is about 0.070 in, as measured from experimental tests recently conducted on unirradiated channels. A permanent set of about 0.015 in results. Such a deformation would not hinder the functional capability of the CRD system. It also should be noted that the highest pressure difference occurs when the break is assumed to occur inside the velocity limiter. The pressure difference caused by a break outside the velocity limiters is 67 percent of the value from a break inside the limiters. 2.7.2.3 Earthquake Loadings The effects of earthquake loadings on the Nine Mile Point core structure were analyzed utilizing the dynamic analyses of the reactor vessel. The resultant loadings were then directly applied to the internals. The results indicated that the NMP Unit 1 UFSAR Section XVI XVI-16 Rev. 25, October 2017 response of the core structure was about 20 percent to 40 percent lower than the reactor vessel, indicating a conservatism in the assumptions. Stresses induced by the design earthquake on the fuel channel and rod guide tubes were 3200 psi and 480 psi, respectively. Deflections were about 0.32 in and less than 0.016 in, respectively, for the above components. The lateral earthquake acceleration force on the core support is directly applied to shear pins that connect the core support to the shroud. Shear stress due to a design earthquake is 6500 psi, which is substantially below the ASME Code allowable for a direct shear application. (Note that the steam line break accident produces negligible loads on these pins.) There are also 4 core plate wedges (spacers) located between the core plate and the shroud inner diameter (ID) that were added with the core shroud repair tie-rod assemblies. The wedges provide a direct load path from the core plate to the shroud ID during an earthquake. The lateral thrust applied to the top guide is sustained by the same pin arrangement as the thrust on the core support. However, backing these pins are eight spacer plates, equally spaced around the outer rim of the top guide. These spacer plates transfer the horizontal loads from the top guide directly to the ID of the shroud. Resultant stresses in the top guide are below ASME Code allowable for the design earthquake. Design earthquake loadings produce stresses less than 1600 psi on the core support and shroud. These stresses and any shroud displacement will not produce any significant effects on the response of the core structure to the differential pressure induced by an accident. 2.7.3 Stresses and Deformations at Which the Component is Unable to Function and Margin of Safety This question is not readily answered, with the possible exception of those members sustaining a buckling load. Consequently, each loading on each component must be individually discussed. Where it was evident that the stress margin of a given component was very large, the margin was calculated using Code allowable stress in the numerator, rather than using loss-of-function stress. Where less margin was evident, the loss-of-function stress was used to calculate the margin. This has the effect of understating the margin whenever the Code allowable stresses are utilized. NMP Unit 1 UFSAR Section XVI XVI-17 Rev. 25, October 2017 2.7.3.1 Recirculation Line Break (See Table XVI-8) Lower Shroud and Core Support Since the limiting applied pressure is external, the limiting applied load is then a buckling (collapse load - about 320 psi). Therefore, when compared with the applied differential pressure, a safety margin of about 2.6 is evident. Upper Shroud If one uses the Code allowable of the material at 550°F as the limiting stress, then the shroud should be able to sustain an internal pressure of over 260 psi, with a resultant safety factor of over 35. Note that a minimum radial stiffness is needed to machine the shroud; thus, the design is not governed by operating stress requirements. Core Support Plate The stress sustained by the core plate during a recirculation line break is not given because a plastic analysis was performed on the structure utilizing an "idealized" stress-strain curve. This stress-strain curve assumes that the stress increases linearly with strain until the yield point of the material is reached, where further strain causes no increase in stress. Further, the analysis utilizes the conservative assumption that the circular plate may be represented by a beam where length is the diameter of the plate. This assumption results in a lower bound solution describing a minimum safety margin. To quote a margin of safety resulting from a conservative analytical model would be misleading. If one were to utilize as the analytical model an equivalent solid plate that is simply supported, then collapse pressures of approximately 700 psi could be obtained. Again, to quote safety factors based on nonconservative models would also be misleading, but in the opposite sense. At what precise pressures ultimate collapse will occur is difficult to compute. Although such computations are not necessary to ensure structural integrity, it is estimated that about 250 psi would be required to ultimately collapse the plate. If such a number were utilized, then a safety factor of 1.9 could be used for this structure. Guide Tube If the ASME Code allowable stress (15,000 psi) were utilized to determine when the component would not function, the margin of NMP Unit 1 UFSAR Section XVI XVI-18 Rev. 25, October 2017 safety would be about 3.7, based on the calculated actual stress of 4300 psi. Fuel Channel The results of an analysis based only on ASME Code allowable stresses can lead to erroneous conclusions because of the small resulting safety margin. However, a collapse of the channel is not a catastrophic occurrence since the inward motion of the channel is limited by the fuel assembly. It is estimated that the fuel assembly will continue to support the fuel channel until pressure in excess of 30 psi is imposed. Use of 30 psi with the 22 psi applied differential pressure yields a safety margin in excess of 1.3. 2.7.3.2 Steam Line Break (See Table XVI-9) Note: Safety factors are quoted utilizing the differential pressures that occur due to a steam line break inside the flow limiter. Lower Shroud and Core Support Assuming the component is unable to function when the stress reaches ASME Section III Code allowable (15,800 psi), the margin of safety is about 4.2. Upper Shroud Making the same assumption as in the Lower Shroud and Core Support above, the margin of safety is about 11.3. Core Support Plate The stress value is given for the bolt subjected to the highest stress. Assuming the structure will fail to function when the highest stress bolt reaches yield, the safety margin is about 2.3. Guide Tube Since failure is by buckling, the absolute limit of the functional performance is an internal pressure of about 106 psi, which results in a safety factor of about 2.6. Fuel Channel NMP Unit 1 UFSAR Section XVI XVI-19 Rev. 25, October 2017 Based on experimental tests, the fuel channel can sustain internal pressure in excess of 30 psi without causing enough deformation to interfere with the scram capability of the CRD; therefore, the safety margin is in excess of 1.3. No significant changes in the mechanical properties of annealed 304 stainless steel are expected to occur until an exposure of 5 x 1020 fast NVT is reached. All the critical austenitic stainless steel structural components receive at least a factor of 2.5 times less than the above amount with the exception of the top guide. (The above factor is conservative because the irradiation exposure analysis model conservatively assumes a flux level attendant with fully withdrawn control rods during the 40-yr design life of the reactor.) The bottom of the egg crate top guide beam comes within 2 in of the top of the active fuel (TAF). The bottom of the beam receives about 3.5 x 1021 fast NVT. However, the beam is over 11 in in depth and the top surface sees only a small fraction of the above exposure. The top guide does not sustain any significant loading from the recirculation and steam line breaks, but does see a structural loading from the lateral acceleration forces induced by an earthquake. In structural applications, annealed 304 stainless steel irradiated to flux levels above 5 x 1020 fast NVT are expected to perform satisfactorily if either of the following conditions are met: 1. The yield strength of the material is not exceeded. 2. The total strain, in both the elastic and plastic region, does not exceed 2 percent where the yield strength is exceeded locally. The above conditions are met for the top guide. Also, since the exposures on the other stainless steel critical core structures are less than 5 x 1020 fast NVT, and since these exposures were calculated on a conservative basis, no detrimental structural effects will occur as a result of radiation damage. The fuel channels are made of Zircaloy; they will receive an irradiation exposure that will affect the material properties after a few weeks of reactor operation. The pressure limits due NMP Unit 1 UFSAR Section XVI XVI-20 Rev. 25, October 2017 to a recirculation or steam line accident, as shown in Tables XVI-8 and XVI-9, are based on the unirradiated stress limits. As a consequence of irradiation, the yield strength more than doubles, and the ductility drops to a range between 0.5 and 1 percent. However, the loads applied to the channel result in a nominal stress less than the yield strength for the irradiated material. Since the channel does not contain any notches or stress risers, it can sustain the pressure differential loads applied during an accident without failure (that is, without affecting the core cage-like configuration or hindering CRD scram capability). 2.8 Safety Margins Against Ductile Fracture Analyses have been performed to demonstrate through fracture mechanics analysis that there exists adequate safety margin against fracture equivalent to those required by Appendix G of the ASME Code for beltline plates having upper shelf energy (USE) values below the 10CFR50 Appendix G minimum value of 50 ft-lbs.(5,6) The Nuclear Regulatory Commission (NRC) issued a safety evaluation dated April 20, 1994, for Unit 1.(7) This evaluation concludes that the Unit 1 RPV plates have adequate margins of safety against fracture until end of life (25 EFPY) for all level conditions (A, B, C, and D), and meet the ASME Section XI Code Case N-512. This conclusion applies to the weld material as well. Further, the Unit 1 reactor vessel plates and weld material satisfy the requirements of 10CFR50 Appendix G, Section IV.A.1, in that the USE values for these plates and welds will provide margins of safety against fracture equivalent to those required in Appendix G of Section III of the ASME Code and are, therefore, acceptable. 3.0 Inspection and Test Report Summary 3.1 Materials The structural material for the vessel is as follows: Head and Shell Plate - SA-302 Grade B Main Closure Flanges - SA-336 Modified Main Nozzle Forgings - SA-336 Modified Nozzle Extensions - SA-336 F8M Drain - SA-105-II NMP Unit 1 UFSAR Section XVI XVI-21 Rev. 25, October 2017 All material used in the fabrication of the pressure vessel was specified by the vendor and approved by GE prior to ordering. A complete set of material specifications is on file at the vendor's plant. The vendor's material certification is given in a report by Combustion Engineering, Inc., Inspection Report for the 213 Inch I.D. Reactor Vessel and Components, CE Serial Nos. 64101 and 64201, CE Contract 164. In compliance with Section I of the ASME Boiler and Pressure Vessel Code, complete mill reports and inspection results are on file with the vendor. A typical mill report is shown in Section XVI-E, Exhibit 1. Typical Charpy tests for the plate material are given in Exhibit 2. The initial NDTT of the weld material is below 10°F in Exhibit 3. A certification of ultrasonic inspection results is included in the above inspection report by the vendor. 3.2 Fabrication and Inspection The vessel, except for the in-core nozzles and the 2-in drain nozzle, was fabricated from rolled and formed carbon steel plate or forgings. All stainless steel cladding was applied by the weld deposit method. The smaller forged nozzles (2 in and below) are Inconel nozzles with partial penetration welds. The CRD nozzles are stainless steel, welded to the vessel with Inconel. The in-core monitors are stainless steel forgings attached to the vessel head by partial penetration welds. The drain nozzle is carbon steel, partially clad, attached to the vessel by full penetration welds. The reactor vessel was fabricated by the Combustion Engineering Company in its Chattanooga, Tennessee, shops during the period of February 1964 to September 1966. All welding was performed by Code-qualified welders and to procedures in conformance with ASME Code, Section IX, which were developed by Combustion Engineering and approved by GE. Each specific weld on the drawing was marked so that the applicable welding procedure used could be identified. The welding procedure specifications for attaching the vessel flange to the vessel shell, as well as many other welds on vessel base material, are shown in Section XVI-E, Exhibits 4 (manual welding) and 5 (machine welding). The welding procedure specifications for applying the stainless steel cladding are shown in Exhibits 6 (manual welding) and 7 (machine welding). NMP Unit 1 UFSAR Section XVI XVI-22 Rev. 25, October 2017 Heat treatment and stress relief of the reactor vessel were in conformance with Section I of the ASME Boiler and Pressure Vessel Code and all applicable Code cases. Preheat and postheat treatment, as required by the welding procedure specifications, were also performed. All weld clad was ultrasonically inspected for bonding to the base material, and all clad material was dye penetrant inspected after final heat treatment. All forgings were inspected by magnetic particle and ultrasonic means to ensure freedom from defects. Imperfections detected in the inspection of the plates, forgings, cladding, and weld were removed and material replaced by the vendor employing welding repair procedures which were approved by GE. During vessel fabrication, multiple cracking transverse to the weld axis was discovered by the welding operator in the weld joining a recirculation outlet nozzle to the lower shell course of the reactor vessel. The weld was approximately three quarters completed when the cracks were found in the weld surface. At this time all shop submerged-arc welding with the materials in question was held up pending establishment of the cause and corrective action. The cause of the cracking was determined to be the particular welding wire, welding flux combination being used. A portion of the final alloy content of the weld deposit was derived from the alloying constituents of the welding flux. It was found that in this instance the wire used and the flux both exceeded the normal range of manganese content. All seams of the vessel, where the use of this combination was possible, were rechecked chemically and ultrasonically. No other cases were found, and later radiographic examination verified the fact. Stricter controls were put into force for qualifying all lots of flux and wire combinations to be used in fabrication, thereby preventing undesirable lots from reaching the floor of the shop. Based upon the results of all inspections, the vessel has received an ASME Code stamp and has been accepted by GE as meeting its specification. All inspections which were required to meet the ASME Boiler Code were witnessed by an inspector licensed by the National Board of NMP Unit 1 UFSAR Section XVI XVI-23 Rev. 25, October 2017 Boiler and Pressure Vessel Inspectors. Continuous inspections were also performed by Combustion Engineering personnel during fabrication. These inspections were audited by GE's vendor quality assurance representatives. Certain inspections, such as dye penetrant inspection of vessel cladding, were witnessed by a GE representative. The certificate of Boiler Shop Inspection along with Combustion Engineering Certifications are given in the vendor inspection report (see Section XVI-3.1 above). The following specifications are included in Section XVI-E. Liquid Penetrant Testing - Exhibit 8 Ultrasonic Testing of - Exhibit 9 Heavy Forgings Ultrasonic Testing of - Exhibit 10 Plate Material Ultrasonic Testing of - Exhibit 11 Weld Overlay Cladding Magnetic Particle Testing - Exhibit 12 Hydrostatic Testing of - Exhibit 13 the Vessel All radiographic testing met or exceeded 2-percent sensitivity. The majority of the radiographs provided 1-percent sensitivity. The limits of acceptance for other testing specifications are shown in Exhibits 8 through 13. 4.0 Surveillance Provisions 4.1 Coupon Surveillance Program The initial nil ductility reference temperature (RTNDT) of the limiting vessel material opposite the core is 36°F. The adjusted reference temperature (ART) increases with increasing fast neutron exposure. Regulatory Guide (RG) 1.99, Revision 2, is used to calculate the ART as a function of neutron exposure. Vessel material surveillance samples are located within the reactor vessel core region to permit periodic monitoring of exposure and changes in material properties. The original, plant-specific material sample program conformed with ASTM E185-66 except for the material withdrawal schedule, which was originally specified in the Technical Specifications. Three surveillance capsules were installed in the Unit 1 reactor in 1969 prior to initial operation. Since plant life extension NMP Unit 1 UFSAR Section XVI XVI-24 Rev. 25, October 2017 is being considered, two capsules (A' and C') were reinserted. The prime indicator is used to designate the new capsule in the same azimuthal location as the original capsules. The radial location of the new capsules is slightly closer to the core than the original capsules to increase the neutron flux. In Reference 40, the NRC approved Unit 1 participation in the BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP), as described in BWRVIP-78 (Reference 37) and BWRVIP-86-A (Reference 38). The NRC approved the ISP for the industry in their safety evaluation dated February 1, 2002 (Reference 39). The ISP meets the requirements of 10CFR50, Appendix H. Participation in the ISP replaces the Unit 1 plant-specific vessel material surveillance program. The current surveillance capsule withdrawal schedule for Unit 1 representative materials is based on the latest NRC-approved version of BWRVIP-86 (Reference 38). No capsules from the Unit 1 vessel are included in the ISP. Capsules from other plants will be removed and specimens will be tested in accordance with the ISP implementation plan. The results from these tests will provide the necessary data to monitor embrittlement of the Unit 1 vessel. BWRVIP Integrated Surveillance Program (ISP) BWRVIP-86A, rev 1-A, Table 4-8 test matrix as deferred. The ISP intention is that plants leave the capsules defined as "deferred" in the reactor design locations such that they continue to receive exposure to preserve ISP contingency planning. veillance capsule) was found dislodged from the vessel pad and was removed from the reactor vessel in April 2015. The BWRVIP ISP has support the ISP. The ISP has also concluded that reinstallation into the reactor vessel is not warranted to support the ISP. The regulatory commitment per letter NMP1L3101, dated September 29, 2016 informed the NRC that NMP1 will maintain the capsule in the ) in the spent fuel pool satisfies the NMP1 License Condition H and the BWRVIP ISP program implementation protocol. 4.2 Periodic Inspection NMP Unit 1 UFSAR Section XVI XVI-25 Rev. 25, October 2017 Periodic inspections of the reactor vessel and its components are performed in accordance with the Inservice Inspection (ISI) Program (Technical Specification 4.2.6). 5.0 Core Shroud Repair Design Description 5.1 Horizontal Weld Repair The reactor core shroud stabilizers are designed to structurally replace horizontal shroud welds H1 through H7. Figure XVI-12a depicts the Unit 1 horizontal shroud welds. The Unit 1 shroud stabilizers consist of two separate design features as shown on Figure XVI-12b. Tie-rod assemblies combined with core plate wedges replace welds H1 through H7 and the upward vertical load-carrying capability of weld H8. The shroud stabilizers are designed to maintain the shroud functions described in Section IV-B.7.0, in the event welds H1 through H7 become cracked 360 deg circumferentially throughwall. The design of the shroud stabilizers is in accordance with the Boiling Water Reactor (8). The shroud repair design was approved by the NRC as documented in NRC Safety Evaluation for NMP1, Evaluation of Core Shroud Stabilizer Design, dated March 31, 1995. Details of the stabilizer design are located in the reference documents listed in Table XVI-9a. Modifications to the tie-rod assemblies were made during refueling outage 14 (RFO14), RFO15, and RFO19 to correct original design deficiencies. Details of the design analyses are included in the shroud repair hardware analysis listed in Table XVI-9a. The original design of the tie-rod assemblies was performed as an alternative to ASME Section XI as permitted by 10CFR50.55a(a)(3), which required NRC approval of the original design. Consequently, the modifications made during RFO14, RFO15, and RFO19 also require approval by the NRC. The NRC safety evaluations that approved the various tie-rod modifications are listed in Table XVI-9a. The NRC safety evaluations describe the design basis of the tie-rod modifications. 5.2 Vertical Weld Repair The reactor core shroud vertical weld repair clamps are designed to structurally replace vertical shroud welds V9 and V10. Figure XVI-12c provides a roll-out drawing of the shroud and depicts the vertical shroud welds. Figure XVI-12d provides a schematic of the vertical weld repair clamps. The vertical weld NMP Unit 1 UFSAR Section XVI XVI-26 Rev. 25, October 2017 repair design was reviewed and approved by the NRC as an alternative code repair pursuant to 10CFR50.55a(a)(3)(i) as documented in References 35 and 36. Core shroud vertical weld repair design documentation is listed in Table XVI-9a. B. PRESSURE SUPPRESSION CONTAINMENT 1.0 Applicability of Formal Codes and Pertinent Certifications The Nine Mile Point pressure suppression containment was fabricated, inspected, and tested in accordance with the ASME Boiler and Pressure Vessel Code, Section III (1965), and all applicable Code cases. The drywell, suppression chamber and connecting vent pipes were designed, erected and tested by the seller, Chicago Bridge & Iron Company. Exhibit 14 of Section XVI-F is the seller's certification that the containment complies with Code requirements. Special precautions not required by Codes were taken in the fabrication of the steel shells. The plate was preheated to a minimum temperature of 200°F prior to welding of all seams thicker than 1 1/4 in, regardless of surrounding air temperature. Preheat at a minimum of 100°F was applied prior to welding of all seams 1 1/4 in thick or less when ambient temperature fell below 40°F. The containment purchase specification required the seller to submit a detailed stress analysis for steady-state and transient conditions. Stone and Webster Engineering Corporation (SWEC) was retained to assist in reviewing the stress report to determine whether or not the report adequately establishes that the design complies with the requirements of the Code. The detailed review of the Code calculations was performed by Teledyne Materials Research Company. This company also prepared an analysis and a stress report for specific piping penetrations of the containment. These reports are included in Section XVI-F of this volume. From the review of the stress report it was determined that the only places that possible deviations from Code existed were where the wording of the Code was ambiguous. The questionable areas were: 1) thermal stress in the sand transition zone, 2) discontinuity stresses at the top flange, and 3) stresses in the expansion bellows. Except for areas listed above, the containment stress meets Code requirements. For the listed areas the Code was ambiguous and interpretation of the Code determines whether the stress report meets Code requirements. NMP Unit 1 UFSAR Section XVI XVI-27 Rev. 25, October 2017 Chicago Bridge & Iron Company initiated a Code case to clarify the design requirements for the questionable areas above. All parties concerned with the review of the containment design concluded that the design was adequate for the intended containment service. 2.0 Design Analysis 2.1 Code Approval Calculations Under Rated Conditions The results of the calculations used to determine if the containment vessels and penetrations meet Code requirements are discussed in Section XVI-F. 2.2 Ultimate Capability Under Accident Conditions To determine the ultimate capability of the containment, it was assumed that a circumferential recirculation line break occurred and all core and containment spray systems failed. The pressure transient was calculated and is shown on Figure XVI-13. The containment is expected to maintain its integrity up to the original test pressure. Suppression chamber integrity could not be ensured after approximately 1300 sec, the time to reach 40.25 psig. The drywell integrity could not be ensured after approximately 55,000 sec, the time to reach 71.4 psig. 2.3 Capability to Withstand Internal Missiles and Jet Forces Several potential missile and jet hazards within the containment were examined. Table XVI-10 lists the potential hazards and data on the forces acting on the shell. Table XVI-11 gives the results of the analysis and demonstrates that none of the jets or missiles considered would cause rupture of the containment. The method used to determine the final deformation of the steel wall and the possibility of penetration is as follows. A load is assumed; the deformation resulting from this load is calculated according to the method of Roark(9) for missiles striking cylindrical portions of the drywell, and according to the method of Bijlaard(10) for spherical portions of the drywell. With the load and deformation known, the strain energy absorbed in the steel can be calculated. For Cases A, B, C, and F the steel wall is backed up by a concrete wall approximately 2 in away. For these cases, the energy required to crush the NMP Unit 1 UFSAR Section XVI XVI-28 Rev. 25, October 2017 concrete is added to the strain energy of the steel to obtain the total energy absorbed. By assuming various loads, a plot can be constructed of the force and energy absorbed in steel and concrete as a function of displacement. The intersection of this curve with a plot of total energy to be absorbed (energy of missile plus energy of jet) yields the value of displacement at which the energy of the missile and jet is absorbed. But this displacement corresponds to a force; if no rupture is to occur, this force cannot exceed the force required to shear the plate. In all cases the energy is absorbed before the displacement force causes the plate to be sheared. 2.4 Flooding Capabilities of the Containment For long-term post-accident recovery, provision is made to flood the containment to above core level. Drywell pressure and water level indication and alarms are provided in the main control room. The stresses on the containment structure resulting from flooding up to el 333' (about 43 ft above the core or 7 ft below the operating floor) have been analyzed. Review of Chicago Bridge & Iron Company computations shows that the containment integrity will be maintained when the containment is flooded to el 333' and subjected to the seismic forces from the design earthquake. Buckling of the shell material will not occur since critical unit loading for buckling is not approached. Table XVI-12 compares the maximum stresses with the critical stresses at which buckling could occur. If containment venting to the atmosphere is not feasible because of high fission product inventory, flooding to above core level cannot be achieved. The containment can be partially flooded, but flooding above el 233' (7.6 ft above the concrete base) could result in a pressure exceeding the original suppression chamber test pressure of 40.25 psig. The pressure suppression system was subjected to a rigorous analysis to determine the maximum plate stresses for a combined temperature and flooding loading condition. The analysis consisted of two parts, the model and the shell. To determine reactions, displacements, and rotations necessary for the shell analysis, the pressure suppression system was modeled as a space frame consisting of 240 joints and 480 members. STRUDEL, the computer program used to solve the model, was developed by the Massachusetts Institute of Technology. The shell analysis was based on the paper, "Numerical Analysis of Unsymmetrical Bending of Shells of Revolution," by Budianski and NMP Unit 1 UFSAR Section XVI XVI-29 Rev. 25, October 2017 Radkowski, using the program, "Unsymmetrical Bending of Shells of Revolution," developed by AVCO Corporation. The stress analysis for combined loads included the following: dead load of the pressure suppression chamber steel, weight of water in flooded pressure suppression chamber, pressure due to the water in the drywell to el 301'-0", thermal (minimum operating = 32°F, maximum post-incident = 205°F), and 0.15g horizontal and 0.055g vertical earthquake accelerations acting simultaneously. The resulting maximum circumferential bending stress at the top of the pressure suppression chamber is 4,870 psi. This stress is then combined with the net circumferential membrane stress of 5,190 psi to give a total maximum stress of 10,060 psi, which is well within normal Code values. 2.5 Drywell Air Gap Thermal and pressure requirements of the drywell determined the size of the air gap. Based on an incident temperature of 310°F and a pressure of 62 psig, the magnitudes of expansion of the drywell shell plate in the critical areas are as indicated on Figure XVI-15. The calculated maximum seismic deflection of the drywell with respect to the reactor building is 0.05 in at the equator of the spherical portion of the drywell. The minimum as-built gap between the exterior of the drywell shell plate and the forms was established at 2 in and was well maintained. The maximum as-built gap from field dimensions is 3 in; the average as-built gap is 2 1/2 in. Shop-fabricated fiberglass forms (1/2 in minimum thickness) with flanges for bolting were used to retain the poured-in-place concrete surrounding the drywell. Detailed form drawings were made based on the containment vessel shop-detailed drawings extrapolated for the 2-in air gap. Individual form panels were from 6 to 8 ft in width and 14 to 16 ft in height. Spacer blocks were wired at the panel corners during erection to ensure proper clearance to the vessel shell. A nominal clearance of 1/2 in between top, bottom, and sides of panels was allowed for adjusting and shimming of the panels. The shimming material consisted of 1/2 in by 3/4 in compressible polyurethane strips. Each panel was bolted in its relaxed state with shims in place and the joints taped with fiberglass mat strip and polyester resin sealant for leak-tightness. After a horizontal course of forms was completed with spacer blocks secure and the gap dimension checked, concrete was poured NMP Unit 1 UFSAR Section XVI XVI-30 Rev. 25, October 2017 in lifts of 3 to 4 ft to about 3 ft below the top of the form to provide free board in protecting the air gap above. The air gap above was closed as forms were erected and before concrete was poured, by stuffing burlap rope into the gap and wiring it to the panels. A plastic sheet was flashed to the vessel above the formwork and draped over the top of the forms to prevent concrete spatter on the vessel shell and the top of forms. The intersections of penetration sleeves and formwork were also sealed with fiberglass strip and polyester resin, as shown on Figure XVI-16. This method produced a tight joint with no leakage problems. Field adjustment and, in some instances, field alteration of the forms was required to meet the specified clearances to the shell plate and penetrations. Field personnel that supervised the air gap dimensioning at the shell plate and penetrations also guarded against foreign objects getting into the air space. All supervisors and craftsmen in this phase of the work were carefully instructed about the necessity of maintaining the air gap in an exact and clear condition. The air gap is ventilated at the top and bottom as follows: at the top of the air gap are two 12-in diameter emergency condenser (EC) pipes just below the head cover of the containment. These pipes, with a 2-in minimum clearance to the penetration sleeves, vent the air space to the outside of the biological shield concrete. At the bottom of the air gap are ten 6 ft 9 in vent pipes that lead to the suppression chamber. These pipes, with a 2-in minimum clearance to the penetration sleeves, vent the air space into that part of the building where the suppression chamber is located. In addition, the annulus between many of the penetrations from the drywell and their surrounding concrete sleeves provides available ventilation area to the air space. The air gap is drained with ten 4-in diameter pipes at the top of the sand cushion as shown on Figure XVI-15. No condensation or water has been found in the drains. 2.5.1 Tests and Inspections Those portions of the air gap that are in the vicinity of the majority of the piping penetrations are available for visual inspection. 2.6 Reactor Shield Wall NMP Unit 1 UFSAR Section XVI XVI-31 Rev. 25, October 2017 The reactor shield wall has been analyzed as a double-walled cylinder with a 21-ft 2 1/2-in ID supported at its base (el 258-1 1/2) and its top (el 303-3). Figure XVI-17 shows the overall plan and structural details. The reactor shield wall is assumed to be a cylinder which is simply supported at its ends and pressurized internally. Both inner and outer 5/8-in thick steel liner plates are stressed in tension due to internal pressure. Since the loading distribution between the inner and outer plates was within 5 percent, they were assumed equal. The plates are connected to the 27WF177 columns by continuous 3/8-in fillet welds (E70 electrodes) on the inner plate and intermittent 5/16-in fillet welds, 6 in long on 12-in centers on the outer plate. These welds are the weakest part of the structure for the internal pressure loading. Allowable stresses as shown in the 1969 AISC Handbook* were increased 50 percent for this loading (see Table XVI-13), which is assumed to be 0.9 of the shear yield stress. The results of this analysis show that the reactor shield wall is capable of withstanding more than 96 psi. The reactor shield wall is a 25-ft, 7 3/4-in outside diameter, circular concrete and steel cylinder attached to the reactor vessel support pedestal and extending upward approximately 45 ft from the base. The inner surface of the annulus is the reactor vessel wall and support skirt. The support pedestal forms the base of the annulus whose top is open to the drywell. The reactor shield wall is 26 2/3 in thick and consists of ten vertical 27WF177 beams, welded together by 5/8-in steel plates, both inside and outside the wall, forming a double-walled shell. The shell is filled with concrete which provides the shielding capability. The top of the shell is covered by a channel-shaped ring girder formed by 1-in thick horizontal steel plate welded to two 1 1/4-in thick vertical steel plates. The recirculation inlet lines go through openings in the wall to the reactor vessel at an elevation near the juncture of the cylindrical portion with the hemispherical lower head. In order to determine the pressure rise inside the reactor shield wall due to a postulated double-ended recirculation line break, it was conservatively assumed that flow from both ends of the broken line discharge directly into the reactor shield Also see Section XVI, Subsection G. NMP Unit 1 UFSAR Section XVI XVI-32 Rev. 25, October 2017 wall/reactor vessel annulus. At full reactor operating pressure the mass input rate would be 50,000 lb/sec, but would decrease rapidly as the reactor is depressurized. However, it was conservatively assumed that the mass input rate to the annulus remained at the high initial value. The annulus vent area consists of an opening of 63 sq ft (assuming that the insulation is crushed or blown away) at the top of the reactor shield wall, ten bottom ventilation holes with a total area of 18 sq ft, and the recirculation line to reactor shield wall clearances which provide an additional 29 sq ft; the total vent area from the annulus is thus 110 sq ft. Moody's critical flow model was used to determine annulus pressure as a function of the break area (6.25 sq ft) to vent area ratio(11). Mixture quality is used as a parameter and is determined by an energy balance. A calculated annulus pressure of 30 psig results if the total available vent area is used. Assuming 15-percent annulus blockage, a pressure of approximately 40 psig is calculated. While the pressure loading will stress the gate support structure to above yield, the stresses still are considerably below ultimate and thus the gate will remain in place. In addition, during the course of the postulated accident, drywell pressure would increase and thereby tend to reduce the differential pressure across the reactor shield wall. Two modes of failure are considered. The first is a guillotine-type break at the pressure vessel recirculation nozzle safe end. The second is a partial failure in which case a piece of the recirculation pipe adjacent to the safe end might break away, allowing a jet to enter the annulus between the reactor vessel and reactor shield wall while the recirculation pipe remains in position. The postulated failure whereby a complete guillotine-type break occurs will cause jet reaction forces acting on both the pipe and safe-end nozzle of 94,750 lb. The reaction force will cause the recirculation pipe to move out of the penetration through the steel shield gates such that the pipe elbow may impact against the containment shell. Due to the shape of the impacting surface, the load would approach that described in Section XVI-B.2.3. No breaching of the containment should occur. Other missiles which could be generated by the broken recirculation pipe will be smaller than the recirculation pipe and have lower forces. The routing of redundant engineered safeguard lines within the drywell has been analyzed to ensure that adequate separation exists so that the rupture and NMP Unit 1 UFSAR Section XVI XVI-33 Rev. 25, October 2017 subsequent movement of one recirculation line will not completely disable any required safeguards system. The remaining jet from the safe-end nozzle consists of water and flashing steam. The water jet is primarily directed through the pipe penetration in the steel gates. However, some flashing steam and water impacts against both upper and lower steel gates with a total force of no more than 67.6 kips, acting radially outward. The second type of failure considered occurs when a part of the recirculation pipe adjacent to the safe-end nozzle breaks away, allowing a jet of water and flashing steam to be directed against the sliding steel gates and reactor shield wall. The recirculation pipe remains in place, effectively blocking the penetration through the steel gates. The maximum radial forces which can be applied to either the upper or lower steel gates are 67.0 kips and 61.8 kips, respectively. The gate guides and their supporting structures will be stressed in bending to a maximum stress of 31,000 psi for this loading condition, which is less than the functional failure stress (0.9 yield stress) of 32,400 psi. Therefore, the steel shielding gates remain in place for either type of failure. Due to its configuration, the reactor shield wall is capable of resisting very large vertical and transverse jet thrusts. The main structural steel members (27WF177) are capable of supporting a radial load of 126 kips at the recirculation outlet using the allowable bending stress of 32,400 psi (0.9 yield stress). Since the jet thrust load is only 94.8 kips, the reactor shield wall can withstand this load. 2.7 Compatibility of Dynamic Deformations Occurring in the Drywell, Torus, and Connecting Vent Pipes Because of the flexible bellows between the vent pipe and the suppression chamber, and the pinned linkage between the header and the suppression chamber, it was assumed that there would be no interaction between the suppression chamber and the vent pipe header chamber for horizontal earthquake motion (see Figure XVI-18). Therefore, these two systems were analyzed separately. In this analysis the drywell was sufficiently stiff to assume the vent pipe header system to be fixed at the drywell. The relative deflections of the two systems at the bellows for the incident and flooded conditions combined with earthquake excitation were found to be well within the allowable design values. NMP Unit 1 UFSAR Section XVI XVI-34 Rev. 25, October 2017 Positive anchorage was provided between the interior concrete structures supporting the reactor and the lower part of the drywell. The anchorage consists of 5 5/8-in by 1-in circular steel shear keys located directly above the support skirt, and welded shear studs on the lower part of the drywell shell in sufficient numbers to resist the sliding motion of the concrete, as shown on Figure XVI-19. The lower part of the drywell shell is, in turn, held to the concrete foundation through the steel supporting skirt welded to the underside of the shell and anchored to the foundation concrete. Concrete is compactly filled in and around the supporting skirt by the preplaced aggregate method so that the skirt and the foundation act in unison. The skirt was not cut following erection. The reactor vessel is held to the drywell shell by a stabilizer system at the top of the shield wall. The drywell shell is, in turn, anchored at the same elevation to the supporting concrete structure. Hence, even though the dynamic characteristics of each of the three elements (the drywell, the reactor vessel, and the foundation) would be different if each were to act independently, they are made to act as a unit by the anchoring arrangements described. Sliding has been prevented by suitable anchoring as described above; there will be no relative displacement between the three structural components. Dynamic analyses were performed as listed below: 1. Reactor support structure, assuming a fixed condition at the base. 2. Drywell containment, assuming a fixed condition at the base. 3. Reactor building. The shears, moments, and relative deflections resulting from these individual analyses were combined to give the acceleration, shear, and moment diagrams for the three structures. These results are shown on Figure XVI-20. Reactor support structure reactions are assumed to be transmitted to the interior concrete slab that supports the reactor pedestal. The inertial forces acting on this interior concrete slab are combined with forces from the reactor support NMP Unit 1 UFSAR Section XVI XVI-35 Rev. 25, October 2017 structure and transmitted to the lower part of the drywell shell. The inertial forces resulting from the dynamic analyses of the drywell shell are combined with the above forces and transmitted to the concrete foundation under the drywell. The anchorages at the interfaces of these three structures were designed to transmit the forces described above without any relative displacement between these interfaces. In addition, loadings due to flooding of the drywell above core elevation were also included in the design requirements for these anchorages. If the bond between the steel and the concrete were broken, the shear force would be carried by the steel keys and shear studs. Since the drywell shell in contact with the concrete was not painted, an epoxy sealing compound (see Figure XVI-21) was placed along the edge of the concrete, both inside and outside the drywell, to prevent water from seeping down along the shell. The sealing compound is sufficiently elastic to expand and fill the gap due to thermal expansion of the steel and concrete. 2.8 Containment Penetrations 2.8.1 Classification of Penetrations Containment penetrations for all piping are divided into five types: 1. Penetrations to accommodate high temperature lines. 2. Penetrations to accommodate occasional high temperature in lines that are cold during normal use, but hot during an incident, and instrument lines that may be heated during periodic blowdown. 3. Penetrations for cold pipes. 4. Control rod drives. 5. Spare penetrations. 2.8.2 Design Bases The design for a typical penetration ensures integrity of all components under a combined normal, seismic, and accidental load by complying with the following requirements: NMP Unit 1 UFSAR Section XVI XVI-36 Rev. 25, October 2017 1. The design meets Code requirements for allowable stresses, and fabrication, test, and inspection methods. 2. The design permits the sealed passage of process piping, instrumentation lines, or CRD pipes through the containment wall. 3. The design accommodates thermal, incident, or combined relative movements between pipes and the containment vessel wall. 4. Provisions for a periodic leak test ensure tightness of a penetration through the entire life of the Station, exceptions being the CRD, ventilation, traversing in-core probe (TIP) system, and spare penetrations. 5. High-temperature penetrations have two expansion joints; the outside expansion bellows are designed for double the design stresses of the internal expansion bellows. The expansion bellows are protected by a guard pipe that surrounds the process line, thus eliminating any direct jet impingement on the bellows and drywell nozzle. In all cases, the anchor near the outside isolation valve for a hot penetration will take jet forces and minimize displacement of the bellows assemblies. 2.8.3 Method of Stress Analysis The containment vessel nozzle extensions were analyzed for stresses due to piping system deflections, vessel movements, seismic effects, thermal transients, and temperature distributions. The metal temperature distributions in the nozzle extensions were calculated by the digital computer program "Lion-L." Stresses due to temperature differences were found by using the "SOR-II" program. All high-temperature piping and high-pressure instrumentation lines were analyzed for thermal stresses by a computer program utilizing vector and matrix algebra. 2.8.4 Leak Test Capability With the exception of the CRD, ventilation, TIP system, and spare penetrations through containment vessel walls, all other NMP Unit 1 UFSAR Section XVI XVI-37 Rev. 25, October 2017 penetrations are equipped with leak-test connections. Testing will be accomplished by pressurizing the double-sealed penetrations. Double expansion bellows for high-temperature penetrations will be tested for leaks together with the remainder of the penetration. Leak detection will then be accomplished by pressure decay techniques. 2.8.5 Fatigue Design Fatigue design is not required for a Class B vessel, as specified by the ASME Code, Section III. Since the penetrations are an extension of the drywell, no fatigue analysis was performed. 2.8.6 Material Specification Seamless carbon steel (ASTM - A106, Grade B) - Main steam and feedwater piping. Seamless stainless steel (ASTM - A376, Type 304) - Other process piping. Carbon steel (ASTM - A212, Grade B (Fire Box) to ASTM - A300) - Containment vessel nozzles and guard pipe. Stainless steel (ASTM - A240, Type 321) - Bellows. To satisfy NDTT requirements, all carbon steel process piping was subjected to a Charpy impact test at -10°F in accordance with Article 12, Section III, of the ASME Nuclear Vessel Code. 2.8.7 Applicable Codes The applicable Codes for the containment penetrations were considered and subdivided in the following manner: 1. The process piping, including valves, was designed, constructed, and tested in accordance with the following: a) Code for Pressure Piping USAS B31.1, including the latest addenda and nuclear interpretations; b) Section I, the ASME Boiler and Pressure Vessel Code and Section III, Article N-324, Non-Destructive Examination, and Articles N-460 to N-469 inclusive, Design of Welded Construction. 2. The nozzles on the containment vessel are an integral part of the containment vessel and are designed in NMP Unit 1 UFSAR Section XVI XVI-38 Rev. 25, October 2017 accordance with ASME Code, Section III, Subsection B. Stresses do not exceed limits stated in paragraph N-1312 (f). 3. All containment nozzle extension piping to the body of the external isolation valve was designed, constructed, and tested in accordance with the Code for Pressure Piping USAS B31.1, including the latest addenda and nuclear interpretations, and with ASME Code, Section III, Subsection B, Articles 11 through 14. The impact tests were conducted in accordance with Article 12 at -10°F. All full-penetration welded joints were fully radiographed in accordance with ASME Section III, Par. N-624, and Section VIII, UW-51. Piping system segments penetrating containment and considered susceptible to thermally-induced overpressurization in the event of an accident were analyzed in accordance with the criteria of the ASME Boiler & Pressure Vessel Code, Section III, Appendix F, 1986 Edition. 4. CRD containment penetrations are designed, fabricated, tested, and inspected in accordance with the rules for Class A vessels, ASME Section III, Summer 1967 addenda, as shown on the detail given in N-462.4 (d) of the Code. 5. The use of expansion bellows in containment vessel penetrations is authorized by Code Case 1330, and falls under the rules of Section III and Code Case 1177. 2.8.8 Jet and Reaction Loads The containment penetrations were analyzed for the effect of a complete line break inside the guard pipes as well as at locations in the drywell vessel up to the connection at the reactor vessel. All calculated axial jet loads inside the guard pipe, with the maximum possible separation of the broken pipe centerlines or with the guard pipe fully pressurized by process pipe breaks, are well below the design strength of the anchors at or near the external isolation valve. Partial process pipe breaks, such as a split inside the guard pipe or complete line breaks inside the drywell, will produce NMP Unit 1 UFSAR Section XVI XVI-39 Rev. 25, October 2017 bending stresses in the process pipe connection to the isolation valve, but the stresses are below the Code allowables. The section moduli of the guard pipe and isolation valves are much larger than that of the process pipe; therefore, their unit stresses are much lower than those in the process pipe. Under all conceivable circumstances the guard pipes will remain intact and prevent jet impingement on the bellows and drywell nozzles. 2.9 Drywell Shear Resistance Capability and Support Skirt Junction Stresses The lower portion of the drywell between the internal concrete and the drywell shell shows a capability to resist shear greater than that required by all load combinations as indicated on Figures XVI-22 and XVI-23. The shear resistance capability is calculated using normal Code design stresses, with no stress increase allowed for any loading condition. If a conservative friction coefficient is assumed, an additional shear resistance can be shown. The shear resistance capability between the drywell shell and the external concrete is also conservative when compared with the allowable shear values indicated on Figures XVI-22 and XVI-23. An additional shear capacity can be shown if a conservative friction factor is assumed. The stresses at the junction of the drywell and its support are shown on Figure XVI-24 for all conditions of loading. The stresses remain below allowable for these loading conditions with no increase in normal Code allowable stresses. 3.0 Inspection and Test Report Summary 3.1 Fabrication and Inspection The drywell, suppression chamber, and connecting vent pipes were fabricated and erected by Chicago Bridge & Iron Company. All welding was performed by Code-qualified welders. All welds were 100-percent radiographed in accordance with Section III (1965) of the ASME Code. Radiographs were graded in accordance with the acceptance standards of the Code. The X-ray units were used with high-speed film and intensifying screens on all main seams. Iridium-192 with fine grain film and lead screens was used on the knuckle seams and flanges. All welds were satisfactory and met the requirements of the Code. NMP Unit 1 UFSAR Section XVI XVI-40 Rev. 25, October 2017 Following construction of the drywell and suppression chamber, each vessel was proof tested. Penetrations were sealed with welded end caps as were the downcomers from the drywell to the suppression chamber. The relief lines from the suppression chamber to the drywell were also blanked. The purpose of the tests was to: 1. Meet the strength specification of 1.15 times design pressure. 2. Meet the leakage specification of 0.1 percent per day. 3. Provide basic data for later tests. 3.2 Tests Conducted In all, five leak rate tests were conducted on the containment vessels. In four of the tests, the drywell and suppression chamber were isolated from each other so that no cross leakage could occur. In the fifth test, the two vessels were interconnected to allow air to transfer between the vessels. All leak rate tests used the reference chamber method of leak testing. Continuous measurements were made of internal and ambient temperatures, barometric and internal pressures, humidity of the vessel air, and differential pressure between the vessels and the reference chambers. Prior to high-pressure testing, each vessel was examined by applying a soap solution over welds and penetrations while maintaining low internal pressure. Leaks found were repaired before the pressure was increased for the strength test. The sequence of testing was as follows: 1. Drywell dry test a. Soapsuds at 5 psig--three leaks repaired, two in control rod penetrations. b. Pressurized to 71.3 psig and held for 1 hr. c. Locks pressurized separately due to pressurizing test line break. d. Soap test at 62 psig--three leaks on end caps of downcomers found and repaired. NMP Unit 1 UFSAR Section XVI XVI-41 Rev. 25, October 2017 e. Leak test for 24 hr. 2. Suppression chamber dry test a. Soapsuds at 5 psig--no leaks. b. Pressurized to 40.25 psig and held for 1 hr. c. Soap test at 35 psig--no leaks. d. Leak test for 24 hr. 3. Suppression chamber wet test a. Soapsuds at 5 psig--no leaks. b. Pressurized to 40.25 psig and held for 1 hr. c. Soap test at 35 psig--no leaks. d. Leak test for 24 hr. 4. Drywell wet test a. Simultaneous with suppression chamber wet test--not connected. b. Pressurized to 37 psig and leak test for 24 hr. 5. Integrated wet test a. Drywell and suppression chamber connected by 4-in line. b. Leak test for 24 hr. The drywell was tested at 62 psig with no water in the vessel (test 1) and at 37 psig with a pool of water at the bottom (test 4). Since the drywell and the suppression chamber pressure were tested simultaneously in the latter case, the drywell was pressurized to 2 psig above the suppression chamber (35 psig) to protect the vent pipes and headers. The pool in the latter test was to provide a water-surface-area equivalent to that which will be exposed to the drywell atmosphere during operation. The suppression chamber was tested at 35 psig without water (test 2) and at 35 psig with water in both vessels and with the air NMP Unit 1 UFSAR Section XVI XVI-42 Rev. 25, October 2017 spaces of both the drywell and suppression chamber interconnected. During the three wet tests, high-precision humidity and temperature instrumentation was installed in the vessels, and circulating fans were used in an attempt to maintain fairly uniform psychrometric conditions over the entire air volume. One humidity instrument and two temperature instruments were placed in the drywell: one for air temperature and one for water. 3.3 Discussion of Results 3.3.1 Results In Table XVI-14, four values for the leak rate are shown for each test. The first value neglects temperature and humidity variations. The second corrects for internal air-temperature changes. The third value is corrected for temperature and humidity changes in the vessel air volumes, and the fourth value is corrected for temperature and humidity changes in both the vessel air volumes and the reference-chamber air. All leak rates are for 24-hr periods with the various parameters averaged for the same time period each day. 3.3.2 Effect of Various Transients 3.3.2.1 Ambient Temperature and Solar Heating of Shell Ambient temperature changes affected the measured leakage from the drywell. In periods of rapidly increasing ambient temperature, the internal air of the drywell increased in temperature nonuniformly. Due to nonuniform air circulation, the air near the metal shell heated up while the air at the drywell center lagged behind. The temperature instrument was at the drywell center, as was the reference chamber. Thus, the drywell average air temperature increased (or decreased) more rapidly than the instrument indicated and more rapidly than the reference chamber. A temperature-induced pressure increase (or decrease) occurred in the drywell, but not immediately in the reference chamber. This showed as a lessening of the differential pressure. A period of 1 to 3 hr after ambient conditions had stabilized passed before the drywell and reference chamber came into temperature equilibrium. Radiant sun heating of the drywell liner had a similar effect. The suppression chamber leakage rate was not significantly affected NMP Unit 1 UFSAR Section XVI XVI-43 Rev. 25, October 2017 by ambient temperature fluctuations since the suppression chamber steel was not exposed. During the drywell wet test on the first day, the ambient temperature increased 10° in 3 hr. This caused a decrease in the differential pressure of 0.37 in H2O, indicating an apparent increase in air mass of 0.026 percent. This is corrected for humidity and temperature, but not for nonuniform psychrometric conditions. In order to account for this nonuniform temperature distribution, numerous temperature instruments at appropriate locations and increased forced air circulation would be required to determine a true average air temperature. Since this instrumentation and equipment was not available, it is necessary to use values for time periods where ambient temperature was relatively stable for 3 hr or more. This condition approximates temperature equilibrium between the drywell and reference chamber. 3.3.2.2 Thermal Lag Through Reference Chamber Wall Temperature lag through the reference chamber wall and air space was checked as a possible source of error. Using results from tests conducted at the Plum Brook Reactor(12), a temperature lag of 0.01°F was assumed for a 2-in diameter reference chamber where the temperature of the air outside the reference chamber changes at the rate of 4°F/hr. In all the tests conducted on the Nine Mile Point containment, this effect caused a negligible error (0.002 percent per day). 3.3.2.3 Condensation in Reference Chamber Changes in reference chamber humidity were accounted for and, in some cases, the error introduced by neglecting this phenomenon was significant. During pressurization, air was introduced into the containment vessel and the corresponding reference chamber simultaneously. Thus, initially both the vessel and the reference chamber had the same humidity and dew point. Then the vessel and reference chamber were isolated psychrometrically by introducing unity oil into the manometer. As ambient temperature decreased, there was condensation on the containment vessel walls, particularly the drywell, even though the average internal air temperature did not get below the dew point. Temperature transients in the air surrounding the reference chamber were slow and temperature lag through the reference chamber wall and the reference chamber air volume were essentially the same. Condensation occurred when the reference NMP Unit 1 UFSAR Section XVI XVI-44 Rev. 25, October 2017 chamber temperature dropped below the initial dew point and should have occurred fairly uniformly through the air volume, not just on the walls as in the case in the drywell itself. As the drywell reference chamber is a vertical tube, any condensation would have tended to drop to the bottom where, due to its small surface area, it was assumed that none revaporized. The difference between the initial reference chamber vapor pressure and the final vapor pressure was added to the final manometer reading. 3.3.2.4 Volume Changes Due to Thermal Transients Temperature changes caused changes in the volumes of the drywell (or suppression chamber) and the reference chamber. Since the thermal coefficient of expansion is greater for the brass reference chamber, its volume changed more drastically. When temperature decreased, the volume decreased. This caused a pressure rise in the contracted vessels. The pressure rise in the reference chamber was greater, thus indicating an increase in the pressure differential between the drywell and reference chamber. The maximum error would be approximately 0.002 percent per day for a 1-deg change. This is significant. When a 10-deg change occurs, the error could be as much as 50 percent of the measured value. These tests demonstrate that substantial errors can be encountered with the testing methods employed. The largest source of error is attributed to the effects induced by temperature transients, which would be far less severe during subsequent retests with the drywell enclosed by the reactor building. The tests also clearly show that humidity corrections can be significant, even without temperature transients, which indicates the need for good internal air recirculation on future tests. 3.3.2.5 Overpressure Test--Plate Stresses The design of the containment is such that, for design basis accident (DBA) loadings and combinations of accident and earthquake loadings, all stresses are within the stresses allowed by the ASME Boiler and Pressure Vessel Code, Section III. During the initial overpressure test of the containment to 1.15 times design pressure, the stresses in containment approached or exceeded the Code allowable stresses. One hundred percent of the plate area had stresses exceeding 71 percent of Code allowable stress. More than one-third of the plate area had stresses which exceeded Code allowable stresses. Table NMP Unit 1 UFSAR Section XVI XVI-45 Rev. 25, October 2017 XVI-15 shows the amount of plate area which had stresses in excess of a certain percentage of Code allowable stresses. Thus, while the overpressure test was not intended to directly simulate accident conditions, the membrane stresses caused by the overpressure test did in fact approximate membrane stresses from accident plus seismic conditions. A comparison of membrane stresses at six locations in the drywell and suppression chamber indicates that, in the mean, the overpressure test membrane stresses were 100 percent of the membrane stresses from accident plus seismic, with a high of 116 percent and a low of 82 percent. C. ENGINEERED SAFEGUARDS 1.0 Seismic Analysis and Stress Report 1.1 Introduction This report describes the original shock analysis to predict the response of the containment spray and the core spray piping arrangements to transient forces generated by the maximum ground motion due to an earthquake at the Nine Mile Point Nuclear Station. Subsequent piping analyses, as a result of modifications, are performed utilizing methods prescribed in ASME Section III-1968. SWEC was requested to compare the results between a static and a dynamic analysis. Static analysis consists of using the static equivalent of the acceleration of the appropriate response spectrum curve. Dynamic analysis uses the actual accelerations for each portion of a system determined from a vibration analysis. At the start of this work, 100 percent of the mass of the piping system was used for the dynamic analysis. As the work progressed, it was established that a correction factor of 41 percent would have to be applied to the piping loads to provide a concentrated load (or mass) equivalent to the uniformly distributed piping load. Curves of the static and dynamic stresses have been superimposed to facilitate comparison. The stresses of these curves reflect the result of thermal expansion, deadweight between supports (where the spans between supports were not limited to those suggested in the ASA piping code), jet spray piping reactions where applicable, and seismic effect. The stresses did not include longitudinal pressure stress or stress intensification factors for screwed fittings or tees; the reason being that NMP Unit 1 UFSAR Section XVI XVI-46 Rev. 25, October 2017 these two stresses are added manually to the computer-generated stresses, and it would be more elucidating to compare stresses caused only by bending and torsion of the pipe. The closing times for the core spray and containment spray valves are 20 and 15 sec, respectively. As the result of calculations, it was determined that these relatively slow closing times have a negligible effect on the stress levels of these systems. Therefore, it was on this basis that water hammer stresses were neglected. The control valves are near the anchor at the concrete wall, outside the containment vessel. Since water hammer shock waves travel upstream of the valve, the spray piping inside the containment vessel (downstream side of the valve) will not be affected by this phenomenon. The shock analysis was a linear-elastic, normal-mode lumped-parameter type using the seismic spectra of the applied transient forces as inputs.(13) For the purposes of the analysis, the spray piping systems were divided into independent piping lines; each line was shock analyzed using the piping flexibility and shock spectrum digital computer programs. Using the piping geometry as input, the piping flexibility program calculated the flexibility. Applying the weight and flexibility of the piping system as input, the shock analysis programs evaluated the normal modal frequencies and the associated modal shapes. Then, using the response spectrum as input, the maximum modal dynamic loads were calculated for each degree-of-freedom. The internal forces, moments, and stresses were calculated, using the maximum modal dynamic loads and the maximum modal displacement and pipe stress analysis program. The modal forces, displacement, and stresses at points of interest were pseudo-statistically combined to obtain the predicted combined-dynamic-response quantities. 1.2 Mathematical Model Each line was mathematically idealized as a three-dimensional elastically-coupled model, with the mass lumped at various points. The analyses were performed using the shock analysis program which makes use of normal mode, linear elastic, and small deflection theories. The flexibilities and stiffnesses of the member were determined by the piping flexibility program and accounted for shear, torsional, axial, and bending deformations and loads. NMP Unit 1 UFSAR Section XVI XVI-47 Rev. 25, October 2017 The materials used in the analyses were assumed to be homogeneous and isotropic and to obey Hooke's Law. Residual stresses in the materials were neglected, but the effect of damping (0.5 percent and 1 percent) for dynamic analysis, and the operating temperature of the system and its effects on material properties were taken into account by using the "hot" modulus of elasticity in the dynamic analysis. Each line was broken into a number of discrete pieces with the mass assumed to be lumped at the midsection of the pieces. The masses of the straight lengths of pipe were based on approximately 5 to 50 ft of pipe length, while the masses of the pipe bends were applied at the center of the bend or close to the bend. The weight of water, 62.4 lb/ft3, was taken into account and combined with the pipe and insulation weight at the respective mass joints. Three dynamic degrees-of-freedom have been allowed at each mass joint (i.e., three translations: X, Y, Z). 1.3 Method of Analysis Normal mode methods for lumped-parameter, linear-elastic systems were used for this analysis. The normal mode method is described in most new textbooks on mechanical vibrations. The techniques used in this analysis are covered in considerable detail in References 14 and 15. Therefore, the following discussion of the method will be limited to a brief review of the basic steps and equations used in the analysis. The dynamic behavior of the actual piping system was approximated in a three-dimensional mathematical model by a finite number of lumped masses which were joined and restrained by weightless structural members. The equations of motion for such discrete systems are in the form of ordinary differential equations and can be readily solved. A shock spectrum type of analysis was used. The following are the basic steps and equations used in the analysis procedure. 1.3.1 Flexibility or Influence Coefficient Matrix The influence coefficient matrix [], as defined here, gives the deflections in the structure for each dynamic degree-of-freedom NMP Unit 1 UFSAR Section XVI XVI-48 Rev. 25, October 2017 due to unit loads at each dynamic degree-of-freedom. The [] matrix is calculated by the static matrix of piping flexibility analysis methods(14,16) using the following matrix equation: [S] [] = [U] Where: [S] = The square stiffness matrix of all mass points of the piping system obtained by combining the stiffness of piping elements(17) [U] = The unit load matrix with unit load vector in the direction of each dynamic degree-of-freedom The calculation of the stiffness matrix of the element includes the axial, bending, shear, and torsional flexibilities. The size of the stiffness matrix for each piping structural element is 6 x 6, since 6 forces and moments and 6 deflections and rotations are considered by the piping flexibility program. 1.3.2 Normal Mode Frequencies and Mode Shapes The modal frequencies and the associated mode shape are calculated by use of the matrix frequency equation: Where: [I] = The diagonal unit matrix, which is comprised of "ones" on the main diagonal and zeros for all of the off-diagonal elements [M] = Diagonal mass matrix [] = Flexibility or influence coefficient matrix The shock spectrum program uses the Wilkinson method and the "cosine rotation theory" to obtain the modal frequencies and mode shapes. The number of frequencies calculated is equal to the number of dynamic degrees-of-freedom assumed in the mathematical model. 1.3.3 The Seismic Spectrum Values NMP Unit 1 UFSAR Section XVI XVI-49 Rev. 25, October 2017 Experience has shown that the behavior of structures during earthquakes is essentially a problem in vibration. The simplest vibrating structure is one having a single degree-of-freedom. Dr. G. W. Housner recommends the use of seismic spectrum response for the evaluation of dynamic loads due to the ground motion. A seismic shock spectrum value gives the maximum absolute response dynamic load in a single degree-of-freedom of particular frequency resulting from application of a seismic force time-history. The force spectrum value fij at a particular degree-of-freedom i and at a natural frequency j for an undamped system may be calculated using: Where: Pi() = The time-history of applied force at the degree-of-freedom Dr. G. W. Housner has determined these response spectra curves by averaging the eight horizontal components of the four strongest ground motions yet recorded and normalized them to an arbitrary ground acceleration. Based upon recommendation of Dames & Moore, the response spectrum curves for ground motion shown in plate C-22, Section III, First Supplement to the Preliminary Hazards Summary Report (PHSR), are used for this work. 1.3.4 Dynamic Modal Loads The maximum dynamic forces in a mode j of a multidegree freedom system can be obtained by use of the shock spectrum values for a one degree system of mode frequency j. The dynamic load vector Fj is obtained by: Where: [W] = The diagonal weight matrix NMP Unit 1 UFSAR Section XVI XVI-50 Rev. 25, October 2017 {j} = Mode shape vector in mode j {j}t = Transpose of the mode shape vector {fj} = Force shock spectrum vector for mode j The dynamic load vectors can be regarded as the equivalent set of static loads. 1.3.5 Modal Response Quantities The displacements, internal forces and moments, and stresses in each significant mode at selected points and cross sections of the piping system are obtained by static matrix analysis, by treating each dynamic load vector Fj as a set of equivalent static loads. The modal joint deflections and rotations are obtained by: Where: [] = The flexibility matrix of all the joints Internal forces and moments in each structural element are obtained as the product of the element stiffness matrix [Sp] and the difference in displacement of the ends of the elements. The stresses in the piping elements are found by use of strength of material formulas and piping code ASA B31.1. 1.3.6 The Combined Response Quantities In order to predict the maximum response quantity (stress, displacement, or force) at a point in the piping system, the modal response values must be combined. Since shock spectrum analysis gives the peak response quantity for each mode, absolute summation of all the modal responses would be too conservative, especially if a large number of modes are considered. Experience with shock test data and analysis of shipboard equipment indicate(18) that the following statistical combination of modal values gives realistic yet conservative results: NMP Unit 1 UFSAR Section XVI XVI-51 Rev. 25, October 2017 Where: Ri = Expected maximum response value at point i Rij = Response at point i in mode j Rim = Maximum modal response at point i a = Number of modes considered This method of combination means that the predicted combined response consists of the two largest modal responses and some portion of the other modal responses. 1.3.7 Basic Criteria for Analysis Piping designs submitted for seismic analysis indicated that many redundant piping systems were similar to each other. Since it was impractical to run analyses on similar systems, engineering judgment dictated the selection of the particular system to be analyzed that would demonstrate the most conservative results. For the thermal analysis, the calculated stresses were held below the allowable stress ranges as indicated in the ASA piping code: ASTM A53 - 18,000 psi ASTM A193 - 5,000 psi ASTM A376 TP304 - 26,250 psi The maximum allowable stress for the seismic plus thermal condition was established at 30,000 psi, the yield point. 1.4 Discussion of Results Figure XVI-25 shows a typical piping isometric diagram with point locations. The highest stressed point for each portion of a system is shown in Table XVI-16. The results demonstrate that no part of either the core spray or containment spray systems is overstressed even during the maximum earthquake. NMP Unit 1 UFSAR Section XVI XVI-52 Rev. 25, October 2017 A comparison between static and dynamic analyses is shown on Figure XVI-26. The stresses shown on these curves do not include longitudinal pressure stress or stress intensification factors for tees or other fittings. These effects are added manually to the computer-generated stresses. The curves compare stresses due only to bending and torsion of the pipe. From Figure XVI-26 and similar curves for the rest of the system, it is demonstrated that the static analysis technique compares favorably with dynamic analysis. 2.0 Containment Spray System The analysis discussed in this section must be supplemented with analyses discussed in Sections XV-C.5.1 and XV-C.5.2. Section XV-C.5.3 specifically discusses analyses applicable to containment spray operability at maximum containment spray raw water temperatures of 84°F. 2.1 Design Adequacy at Rated Conditions 2.1.1 General The purpose of the containment spray system is to condense steam in the pressure suppression system and remove heat from the system. 2.1.2 Condensation and Heat Removal Mechanisms Water is pumped from the pressure suppression pool through a heat exchanger to the spray nozzles within the containment vessels. The water breaks up into droplets as it sprays into the drywell and suppression chamber. Heat is transferred to the spray droplets mainly by two mechanisms--convection heat transfer and mass transfer. The former represents the surface heat transfer from the hot steam-gas mixture to the droplet and is represented by: Qc = hA (Tm - Td) (1) Where: Qc = Rate of convective heat transfer from the mixture to the droplet surface, (Btu/hr) h = Convective heat transfer coefficient, (Btu)/(hr)(ft2)(F) NMP Unit 1 UFSAR Section XVI XVI-53 Rev. 25, October 2017 A = Surface area of the droplet, (ft2) Tm = Temperature of the steam-gas mixture, (F) Td = Temperature of the water droplet surface, (F) The mass transfer represents the latent heat of the steam condensed on the surface of the water droplet and is represented by: (2) Where: Qmt = Rate of heat transfer due to condensing steam, (Btu)/(hr) Kg = Mass transfer coefficient from liquid interface to the gas, (lbm)/(hr)(ft2) (atmosphere) = Heat of condensation of water, (Btu)/(lbm) A = Surface area of the droplet, (ft2) pps = Partial pressure of steam, (atmosphere) ppd = Partial pressure of water vapor at droplet temperature, (atmosphere) Combining both forms of heat transfer: (3) Since partial vapor pressures are a function of temperature, over a relatively small range: (4) Where: Qt = Rate of heat transfer, (Btu)/(hr) Us = Overall surface heat transfer coefficient for convection and mass transfer, (Btu)/(hr)(ft2)(F) Heat flows from the surface of the droplet inward by conduction and is represented by the partial differential equation: NMP Unit 1 UFSAR Section XVI XVI-54 Rev. 25, October 2017 (5) Where: = k/CF = Thermal diffusivity, (ft2)/(hr) k = Thermal conductivity, (Btu)/(hr)(ft2)(F/ft) Cp = Specific heat, (Btu)/(lbm) (F) = Density, (lbm)/(ft3) r = Drop radius, (ft) = Time, (hr) The surface temperature of the drop is assumed to immediately take up a temperature close to that of the steam-gas mixture and remains constant. The mean temperature of the drop is represented by: (6) Where: Ta = Mean temperature of the drop, (F) Ti = Initial temperature of the drop, (F) d = Drop diameter (ft) The mean temperature difference between the droplet surface temperature (Td) and the mean drop temperature (Ta) is represented by: (7) NMP Unit 1 UFSAR Section XVI XVI-55 Rev. 25, October 2017 Heat transferred to the droplet from the steam-gas mixture goes to heating the droplet: (8) Rearranging equation (4) in terms of the mean temperature of the drop: (9) Where: H = Overall heat transfer coefficient for convection and mass transfer, (Btu)/(hr)(f2) [Note: H U] As a close approximation it is assumed that the droplet surface temperature in equation (9) is the same as the mixture temperature: (10) Equations (8) and (10) may be rewritten: (11) or Equations (6) and (7) are used to rewrite (11): (12) Values of are plotted on Figure XVI-27. The Nusselt group approaches a value of 10 as (Ta - Ti/Td - Ti) approaches unity. With k = 0.38 (Btu)/(ft)(hr)(F) for water, the mean heat transfer coefficients can be calculated as a function of drop diameters and are shown in Table XVI-17. NMP Unit 1 UFSAR Section XVI XVI-56 Rev. 25, October 2017 The validity of the assumption that the droplet surface temperature rapidly takes up the mixture temperature [Equations (9) and (10)] can be demonstrated by the kinetic theory of gases using Clapeyron's equation: (13) Where: Us = Surface heat transfer coefficient, (Btu)/(ft2)(sec)(F) fe = Dimensionless factor M = Molecular weight of the vapor, 18(lbm)/(lb-mole) (lbf) (ft) R = Gas constant, 1545 (lb-mole) (°R) = Latent heat of vaporization, (Btu)/(lbm) J = Mechanical equivalent of heat, 778(ft)(lbf)/(Btu) vs = Specific volume of steam, (ft3)/(lbm) vw = Specific volume of water, (ft3)/(lbm) lbf = (slug)(ft/sec2) = (32.2)(lbm)(ft/sec2) The dimensionless factor fe depends on pressure and allows for the effect of molecular collisions between the vapor and the surface. Values range from 0.036 to 1.0. Even using the lowest value, the surface coefficients are high compared to the overall coefficients as shown in Table XVI-18. Using the relationship of equation (9) and the values in Tables XVI-17 and XVI-18, it is shown that the difference Tm - Td is small compared to Tm - Ta. Experimental results(19) have verified the values of the heat transfer coefficient given in Table XVI-17. The heat transferred from the steam to the drop depends on the time the drop remains in the steam. This is a function of the droplet velocity and travel distance. When both loops of a system are in operation, the droplet nozzle exit velocity is 95 NMP Unit 1 UFSAR Section XVI XVI-57 Rev. 25, October 2017 fps. For one-loop operation, the nozzle exit velocity is 56 fps. This velocity is reduced by the effect of the drag force acting on the drop as it passes through the steam-gas mixture. The drag force is given by(20): F = C (1/2 d)2 (1/2) g V2 (14) This force decelerates the drop and is represented by: F = ma = (4/3)()(1/2 d3) w dv/d (15) Where: F = Force due to drag, (poundals) C = Dimensionless drag coefficient g = Density of mixture, (lbm)/(ft3) w = Density of drop, (lbm)/(ft3) V = Velocity, (ft)/(sec) Combining equations (14) and (15) and integrating gives: (16) Where: s = Distance traveled by the drop, (ft) K = (3/4) g/w (D/d) 2.1.3 Mechanical Design Each containment spray system is designed to deliver 5600 gpm to the drywell and 400 gpm to the suppression chamber with both loops of the system in operation. Most of the drywell flow is through 1/2-in diameter nozzles located on ring headers inside the lower spherical portion of the drywell. The design pressure at each of the two supply points is 75 psig. Calculations show that at peak containment pressure (35 psig), a pressure of 85 psig in the primary sparger header and 89 psig in the secondary sparger header can be achieved. NMP Unit 1 UFSAR Section XVI XVI-58 Rev. 25, October 2017 As verified by tests described in paragraph 2.2 below, the droplet size for two-loop operation and a pressure drop of 73 psig is no more than 500 microns. The droplet size for one-pump operation is no more than 1000 microns. The spray nozzles are directed so that droplets do not contact any surface within 5 to 6 ft. Using Equation (16), the droplets would contact a surface in 0.074 sec for the two-pump case and in 0.11 sec for the one-pump case. Inserting these times in Equation (6) shows that the drop temperature rise is 91 percent of the maximum for the 500-micron drop and 70 percent of the maximum for the 1000-micron drop. For conservatism it is assumed that heat transfer ceases once the drop strikes a surface. In reality, most drops would not strike a surface for many more feet and would continue to absorb heat even after contact. 2.1.4 Loss-of-Coolant Accident Following blowdown of steam and water to the drywell and subsequent purging of steam and nitrogen to the suppression chamber, the pressure in both chambers equilibrates at about 21 or 22 psig. This assumes that all the drywell nitrogen is purged to the suppression chamber. As steam purges it is condensed in the suppression chamber pool. Following this initial transient, the only mechanisms for increasing containment pressure are the energy inputs due to decay heat and metal-water reactions. Heat exchangers in the containment spray loops remove heat from the containment and transfer it to raw cooling water. Assuming that all the energy released in the reactor core is used to produce steam results in the highest possible pressure conditions in the containment. During the metal-water reaction, hydrogen is released to the drywell and subsequently purged to the suppression chamber. In order to prevent this purging of hydrogen and possible overpressurization of the suppression chamber, the containment sprays condense steam in the drywell and reduce the steam pressure. This reduces the driving force needed to purge the hydrogen. The spray water enters the heat exchangers at the temperature of the suppression chamber. Each of the four heat exchangers is sized to remove 60 million Btu/hr when the spray flow and raw water cooling flow are each 3000 gpm. This capacity is based on reducing spray flow temperature from 140°F to 100°F for the maximum cooling water temperature of 77°F. NMP Unit 1 UFSAR Section XVI XVI-59 Rev. 25, October 2017 The Section XV-C.5.3 design basis reconstitution suppression chamber heatup analysis verified that the containment design basis heat removal requirements are satisfied at the maximum containment spray raw water temperature of 84°F. Using metal-water reaction rates from Section XV, the pressure transient in the containment is calculated with the results shown on Figure XVI-28. The pressures were calculated for four cases: 1. Core spray is inoperative and only one containment spray pump operates. 2. Core spray is inoperative and two containment spray pumps operate. 3. One core spray pump and one containment spray pump operate. 4. One core spray pump and two containment sprays operate. For cases 1 and 2 there is a 27-percent metal-water reaction. Less than 1-percent metal-water reaction occurs in cases 3 and 4. The previous cases were analyzed for initial containment pressure of 0 psig and suppression chamber temperature of 90°F. As a limiting case it is assumed that core spray is inoperative and only one containment spray pump operates. The results are shown on Figure XVI-29. Note that the suppression chamber pressure reaches the design value after about 1000 sec. 2.2 Summary of Test Results 2.2.1 Spray Tests Conducted A sample spray nozzle of the size and type used in the containment spray system was tested at the Huntley Station in Buffalo, New York. Water was run through the nozzle at various pressures from 10 psig to 100 psig and spray pattern and spray particle fineness observed. A close-up of the nozzle and spray pattern for 80 psig pressure drop is shown on Figure XVI-30. The spray pattern is more clearly defined on Figure XVI-31. This shows that the spray NMP Unit 1 UFSAR Section XVI XVI-60 Rev. 25, October 2017 breaks up into a misty rain with gravity having little effect on the small droplets. Figure XVI-32 is a close-up of the nozzle and spray pattern for a 30 psig pressure drop. The spray pattern for this pressure is shown on Figure XVI-33. This shows that the spray breaks up into a moderate rain with gravity causing the particles to fall as opposed to the many suspended particles on Figure XVI-31. The above nozzle pressure drops of 80 psig and 30 psig represent the pressure conditions for two-pump and one-pump operation, respectively. Table XVI-19 shows the relationship between particle size and the type of spray pattern. The particle sizes for two-pump operation are in the range of 10 to 400 microns. For one-pump operation, particle sizes range from 500 to 1000 microns. The nozzle used in the test had a throat size of 0.6-in diameter. The largest nozzles used in the containment spray system are 0.5-in diameter. This smaller nozzle size results in greater breakup of the spray than was actually observed. The containment spray system as designed should result in smaller droplet sizes than assumed in Section C.2.1. 3.0 Core Spray and Containment Spray Suction Strainers The installation of new core spray and containment spray suction strainers under Modification N1-96-005 involved adding piping and components inside the torus. The required new hydrodynamic load generation and piping system analysis followed the existing methodologies documented in the Plant Unique Analysis Report (PUAR). The PUAR is a plant-specific Teledyne document providing a summary of the analytical methods used and stress results obtained during the original Unit 1 Mark I analysis performed by Teledyne. Under Modification N1-96-005, a PUAR supplement has been prepared and issued under Safety Evaluation 98-104. D. DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 1.0 Classification and Seismic Criteria Class I Structures and Components Structures and components whose failure could cause significant release of radioactivity NMP Unit 1 UFSAR Section XVI XVI-61 Rev. 25, October 2017 or which are vital to safe shutdown and isolation of the reactor. Class II Structures and Components Structures and components which are important to reactor operation but are not essential to safe shutdown or isolation, and whose failure could not result in substantial release of radioactive materials. Class III Structures and Components Structures and components that are not essential for safe shutdown and isolation of the reactor and whose failure will not result in significant release of radioactive materials. No quantitative basis was used to determine the limit for significant release of radioactivity. The basis used was that if a system could fail such that the failure could result in a continuous, uncontrolled release of radioactivity that could not be readily terminated, the system was designated as Class I. Thus, since release from a broken main steam line (MSL) can be terminated by closing the automatic isolation valves, the parts of the system outside the isolation valves are not Class I. Similarly, rupture of a tank in the waste disposal building could result in a release not easily controlled or terminated. Therefore, these systems are Class I. The decisions as to whether the balance of systems, components, and structures qualified for treatment as Class II or Class III were ultimately based on the best professional engineering judgment of those involved in specifying the design criteria. The prime consideration in deciding on Class II or Class III was whether or not the given system, component, or structure is necessary to continued Station operation. Where doubt existed as to which criterion should be applied, generally the resolution was made in the conservative direction, namely, to apply the Class II criterion. A list of Class I and Class II structures and systems is provided below: Class I Structures Reactor building Waste disposal building Ventilation stack Drywell Reactor pressure vessel and its support structure NMP Unit 1 UFSAR Section XVI XVI-62 Rev. 25, October 2017 Suppression chamber Diesel generator support foundation Offgas building Radwaste solidification and storage building Class II Structures Turbine building Turbine generator support foundation Intake and discharge tunnels Administration building extension Combination Class I and Class II Structure Screen and pump house Class I Equipment, Systems, or Areas in Class II Structures Diesel generator support structure Control room Auxiliary control room Battery room Battery board room Supporting steel structure for emergency condenser, makeup, and demineralized water tanks Diesel generator board room Class I Piping Systems Main steam inside drywell Core spray Combustible gas control Containment spray Containment spray cooling water Emergency cooling Liquid poison Drywell and suppression chamber vacuum relief Fuel pool cooling and filtering Reactor cleanup Reactor shutdown cooling Reactor head spray Condensate storage Condensate pump suction and discharge Feedwater booster discharge High-pressure reactor feedwater Reactor building closed loop cooling Control rod drive piping NMP Unit 1 UFSAR Section XVI XVI-63 Rev. 25, October 2017 Radioactive waste disposal system Emergency ventilation Breathing air Instrument air Emergency service water Diesel generator fuel oil, starting air, and cooling water Offgas Reactor recirculation Drywell sump piping (drywell sump to external isolation valve) Class II Piping Systems Main steam outside drywell Bypass steam to condenser Steam supply to air ejector Extraction steam piping Makeup demineralizer Turbine building closed loop cooling Reactor and turbine buildings, sump pump discharge Seal water Turbine oil storage City water Laboratory drains Offgas (turbine gland seal exhaust) Class I Equipment Housed in and Supported by Combination Class I and II Structures Emergency service water pumps and piping Containment spray cooling pumps and piping Diesel generator cooling water pumps and piping Class I Equipment Housed in and Supported by Class II Structures Condensate storage tanks and piping Condensate pumps, suction and discharge piping Feedwater booster pumps and discharge piping High-pressure reactor feed pumps and discharge piping Diesel generator fuel oil, starting air and cooling water piping Emergency condenser storage tanks Reactor building closed loop cooling piping (partial) Breathing air piping (partial) NMP Unit 1 UFSAR Section XVI XVI-64 Rev. 25, October 2017 Instrument air compressors Instrument air piping (partial) Emergency service water piping (partial) 1.1 Design Techniques 1.1.1 Structures The design basis load combinations of dead load, live load (including piping, equipment, and temperature), moving loads, and incident loads are directly combined with horizontal and vertical earthquake loads for structures consisting in whole or in part of Class I elements. The resulting stress levels are within normal Code* values with no increase allowed for the earthquake condition for Class I structures or components except for: 1. Suppression chamber columns, and 2. Ventilation stack for which a one-third increase was allowed. Tables XVI-20 through XVI-26 present the load combinations and allowable stresses for structures consisting in whole or in part of Class I elements. Figures XVI-34 through XVI-41 present the computed deflections from the design earthquake excitation. For concrete design criteria such as bar spacing, bar cover, minimum reinforcement, temperature steel, etc., ACI Code 318-63* was used. For proportioning of concrete members, Part IV-A, "Working Stress Design," of Code 318-63 was followed. The reinforced concrete ventilation stack was analyzed and designed in accordance with ACI Code 505-54. The AISC Specification* for the design, fabrication and erection of structural steel for buildings was rigorously followed in the analysis and design of all structural steel framing and components. For structural components not covered by this specification, applicable documents such as the Uniform Building Code (UBC) and manufacturer-referred specifications were used(21). Also see Section XVI, Subsection G. Also see Section XVI, Subsection G. NMP Unit 1 UFSAR Section XVI XVI-65 Rev. 25, October 2017 It should be noted that Unit 1 has been historically conservative in the application of design specifications. A large-scale static structural analysis was conducted with internally developed two- and three-dimensional-matrix structural analysis computer programs. These programs utilize the stiffness method and are similar to programs such as "STRESS" and "FRAN." Hand computations for static structural analysis utilized classical techniques such as moment distribution, slope deflection and energy methods. The dynamic analysis of each Class I and Class II structure was conducted in the following manner. The structure was idealized as a multilumped mass system interconnected by weightless structural elements. These structural elements took into consideration flexural, shear, and axial deformations. The moment of inertia for a particular story was calculated directly from the plan view of that story. The shear stiffness of each significant structural component was determined individually, then summed directly to give the total story shear stiffness. Utilizing the story moment of inertia and the total story shear stiffness, an equivalent fictitious shear area was calculated. In this manner the structural element representing the story has the same flexural and shear characteristics as the building story. Rock-structure interaction was considered by placing an equivalent spring system between the building and the ground. The method of determining the stiffness of the equivalent spring system was based on the work of Whitman and Richart(22). The weight concentrated at each floor level consists of floor weight, one-half of walls and columns above and below, tanks, piping and equipment associated with the floor, and 25 percent of the design floor live load. The design floor live load was extremely high to satisfy equipment laydown requirements; supplying the full design floor live load would be unrealistic. The mathematical model developed for the reactor building is shown on Figure XVI-42. The dynamic analysis was conducted with an internally developed computer program. The stiffness method of matrix structural analysis(23) determined the deflection influence coefficient matrix. This matrix was postmultiplied by the mass matrix(24). The frequencies and mode shapes for at least the first five modes were determined by interaction with the dynamic matrix(24). The use of the mode shapes to effect a transformation into normal coordinates uncoupled the equations of motion. These NMP Unit 1 UFSAR Section XVI XVI-66 Rev. 25, October 2017 equations were then solved individually by Laplace transforms and the velocity spectrum curves(25), as specified in the Preliminary Hazards Summary Report (PHSR), to obtain the maximum accelerations for each mode. The individual modal accelerations were then combined on a root-mean-square basis(26) to obtain acceleration curves for the structure. The mass at each node point was multiplied by the acceleration at that point to obtain a set of static forces acting on the structure. These static forces were then used to obtain shears, moments, and deflections for the building model. The center of mass was calculated for each floor. The center of rigidity and rotational stiffness were determined for each story. Torsion effects were introduced in each story by applying a rotational moment about the center of rigidity of the story under consideration. The rotational moment was calculated as the sum of the inertial force applied at the center of mass of each floor above and a moment arm equal to the distance from the center of mass of the floor to the center of rigidity of the story. To be conservative, the absolute values of these moments were used in the sum. The torsional moment and story shear were distributed to the resisting walls and columns in proportion to each individual stiffness(27). The normal forces acting on the individual walls and columns were calculated from the story bending moment using beam flexure theory. The following discussion illustrates how the interaction of interconnected structures with dissimilar response characteristics was taken into consideration. The reactor and supporting structure were subjected to a dynamic analysis assuming that the motion of the reactor building at the upper stabilizer level was the same as the motion at the base of the support structure. From the dynamic analysis of the reactor building, the building displacement at the upper stabilizer was calculated as 0.072 in. The upper stabilizer attachment point was displaced 0.072 in and the resulting shears, moments, and deflections in the reactor and support structure were determined. These shears, moments, and deflections were then added to the shears, moments, and deflections resulting from the dynamic analysis to obtain the design values for these quantities. This combination is illustrated on Figures XVI-43 through XVI-45. 1.1.2 Systems and Components NMP Unit 1 UFSAR Section XVI XVI-67 Rev. 25, October 2017 The design basis loads are as follows. Normal loads consist of design pressure and temperature along with dead load. Dead load is that load imposed upon the equipment due to the weight of fluid, piping, etc. Accident loads result from external pressure and temperature conditions to which the system could be subjected. Although the effects of these loads were found to be minimal in the majority of cases, they were considered in the design of all mechanical systems. For isolation valves and anchors on process lines connected directly to the reactor, jet forces were also considered as part of the accident analysis. Water hammer loads are those due to the instantaneous stoppage of the fluid flow. Seismic stresses and stresses caused by operating and other loads were added directly to ensure a conservative design for all structures. Wind and seismic loading conditions were not considered as acting together. For mechanical components, seismic loads were specified as a percentage of gravitational load depending on the physical location of the equipment within the structures. In the design of the piping systems, a "method of differences" approach was used. The calculated stresses produced by thermal action, anchor displacement, and by weight were kept low--under 90 percent of normal design Code allowable. Therefore, the difference to Code allowable and the 20 percent over Code allowable (1 percent of operating time) could be used for any seismic stresses that might be incurred. The loads are combined as follows (Table XVI-27): I Normal and seismic loads are combined for components in accordance with the purchase specifications. For piping systems, the seismic load is added to the normal load by the method of differences, as stated above. II Normal, seismic, and accident loads are combined, except that jet forces due to complete line breaks are not added to normal or seismic loads. III Normal internal and water hammer load stresses are calculated separately, then added together. However, these stresses were not added to external loadings NMP Unit 1 UFSAR Section XVI XVI-68 Rev. 25, October 2017 (such as thermal expansion or seismic loads) as allowed by Piping Code ASA B31.1. Allowable Code stresses, rather than deformation, were used as the governing factor in design criteria. All systems remain within Code allowable stresses when subjected to the load combinations delineated in Table XVI-27. 1.2 Pipe Supports The piping systems were statically analyzed to design seismic hangers and control devices. The design acceleration values were 0.37g horizontal and 0.185g vertical. Forces on the seismic control devices were determined by calculating the pipe load that would be exerted on the device and multiplying the load by the seismic acceleration factor. Hangers and supports were located in accordance with Section 6 of ASA B31.1-1955. Where possible, the vertical seismic loading was incorporated in the design of the vertical hanger to simplify the hanger fabrication. This was possible where rigid vertical hangers were used on cold systems. The location of the additional vertical and all horizontal seismic control devices was determined in accordance with the following. Each individual pipe run of appreciable length was treated separately, and devices were designed to limit movement to an acceptable value (usually 1/8 in +/- 1/16 in clearance, except for those situations that require additional tolerance for thermal expansion). A seismic loading along the longitudinal axis of the pipe run is stopped by one device with the assumption that the pipe acts as a column. The location of devices to stop transverse motion of the pipe run is dependent on the length of the run and has been made to keep bending stresses within allowable values. Guides, stops, or tension-compression struts were used as seismic control devices on systems where the thermal expansion did not produce stresses above 80 to 90 percent of Code allowable. Where additional displacement due to seismic excitation would produce stress levels above 80 to 90 percent of Code allowable, hydraulic snubbers were used. These snubbers allow gradual movement due to thermal growth, but act as a solid member if subjected to a sudden displacement. Snubbers of this type were used where necessary to meet the above criteria. NMP Unit 1 UFSAR Section XVI XVI-69 Rev. 25, October 2017 A testing, inspection, and repair program has been performed in accordance with Inspection and Enforcement Bulletin 79-02 on pipe support base plates and concrete expansion anchor bolts(28). Snubber inservice testing and examination is performed in accordance with Subsection ISTD of the ASME OM Code except where relief/alternatives have been approved in accordance with 10CFR50.55a. 1.3 Seismic Exposure Assumptions The seismic exposure assumptions were presented and discussed in Volume II, Appendix C, of the PHSR. These assumptions are still considered valid and applicable, and are the basis for the selection of the seismic design criteria. Additional studies after the original report, including consultation with personnel of Lamont Geological Laboratory and the New York State Museum and Science Service, have substantiated the original conclusions. The formulae of Kanai for basement rock motions, referred to in the original report, have recently been revised to include data from many small close earthquakes; however, since the controlling earthquake for this project is relatively distant, the original Kanai formulae are generally appropriate. Amplification spectra were developed using the theoretical approach of Kanai. This approach is based on available theory of elastic wave propagation (it is valid in this context to assume that competent rock behaves like an elastic material) and has been verified by comparison with observational data from numerous earthquakes. Ground motion spectra were developed from the basement particle motion and the amplification spectrum. Response spectra were developed from the ground motion spectrum using a structural magnification factor of four (for the undamped spectrum), as has been suggested by G. W. Housner(29). Response curves for various degrees of structural damping were developed using factors proportional to the damped spectra computed for the Helena, Montana, earthquake. These curves were presented in the First Supplement to the PHSR in response to Question III-1 (c). The damping factors used in the dynamic analysis of both Class I and Class II structures were those specified in the First Supplement to the PHSR in answer to Question III-1 (d), with the NMP Unit 1 UFSAR Section XVI XVI-70 Rev. 25, October 2017 exception of the value specified for reinforced concrete structures (integral). In the interim between submittal of the PHSR and the final analysis of structures, published research studies indicated that, in view of the low stress levels in most structures, a damping factor of 5 percent of critical would be more appropriate for integral reinforced concrete structures, with the exception of the ventilation stack. The relatively high stress levels in the stack justify the use of a damping factor of 7 1/2 percent of critical for that structure. There are no reinforced concrete framed structures in the Station. The original design and construction code is specified for each system described herein. Modifications, repairs, replacements, and other work are performed in accordance with the original codes, or in accordance with later editions of those codes, or other codes which have been reconciled in accordance with the requirements of ASME XI. This code reconciliation ensures that structures, systems, and components are designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the function to be performed and commensurate with the original code requirements. 2.0 Plant Design for Protection Against Postulated Piping Failures in High-Energy Lines Unit 1 was designed and constructed prior to 10CFR50 Appendix A General Design Criteria (GDC) 4 and was not designed in accordance with this criteria. The original design basis for Unit 1 is that the probability of double-ended guillotine pipe rupture is extremely low such that protection from the dynamic effects of that rupture were not considered. The licensing basis is that the inherent features and capabilities provide a basis for reasonable assurance that the facility design meets the intent of the criteria. In this regard, pipe whip coping analyses were performed which concluded that containment integrity was maintained with no loss of function, and that the engineered safeguard systems provide core cooling and safe shutdown capability. These analyses utilized functional criteria equivalent to "best estimate" evaluations which are consistent with analyses of beyond design basis low probability hypothetical events. 2.1 Inside Primary Containment NMP Unit 1 UFSAR Section XVI XVI-71 Rev. 25, October 2017 All high-energy lines inside the primary containment have been analyzed for the effects of pipe whip. Table XVI-28 lists all the high-energy systems inside the primary containment. It was assumed that any one of these lines could break anywhere inside the primary containment. A separate analysis was done on the effect of a pipe break on the remaining systems inside primary containment, the effect on safe shutdown capability, and the effect of impacting the primary containment wall. 2.1.1 Containment Integrity Analysis A review of the piping systems inside the drywell was made to determine which systems could, if postulated to fail, impact the containment with sufficient energy to cause concern. The three systems considered as containing sufficient energy upon impact to represent the worst cases are: 1. Reactor recirculation. 2. Main steam. 3. Feedwater. The fluid forces generated by the break were calculated and are discussed below. These loads were then applied to a model of the pipe and the impact velocity at the containment was obtained. The stresses in the containment were calculated by modeling the containment, air gap, concrete, and impacting missile with the appropriate mass and velocity. 2.1.1.1 Fluid Forces Recirculation Loop The force at the break rises from a value equal to the product of pressure times area to about 1.125 times that value in a short time, then decays slowly. However, due to the break in the RPV nozzle, some conservatism was added to cover possible impingement effects of the jet escaping from the nozzle. The actual load applied to the piping system was 1.5 times the product of pressure and area. Main Steam NMP Unit 1 UFSAR Section XVI XVI-72 Rev. 25, October 2017 The force at the break rises from an initial value equal to pressure times pipe cross-sectional area to a maximum of 1.26 times that value in 0.052 sec. Feedwater The force at the break rises from an initial value to a maximum of 1.125 times that value in 0.068 sec. Jet Impingement Force As a result of a break in the recirculation system at the vessel nozzle, a jet is generated which can impinge on the containment vessel. The jet pressure is 40 psi over an area of 26,300 sq in. 2.1.1.2 Impact Velocities and Effects The velocities of the pipes and beams which were analyzed for impact on the containment vessel are given below: Impact Velocity System (ft/sec) Recirculation 115 Main Steam 100 Feedwater 89 Structural Beam (10WF33) 115 The break in the main steam system represents a break at the vessel nozzle. For a break near the main steam penetration, the steam line cannot reach the containment vessel without striking the feedwater system. This would reduce the velocity of the steam line significantly. Assuming that the steam line did not strike the feedwater line, an impact velocity of 270 fps would result. The allowable strain in the containment vessel demonstrates the adequacy of the structure against impacts resulting from pipe break and associated whip. The ultimate strain of the containment vessel is 10 percent. The calculated accumulative strains for the conditions analyzed are given below: NMP Unit 1 UFSAR Section XVI XVI-73 Rev. 25, October 2017 Vessel Thickness Strain Condition (in) (%) Recirculation Loop Elbow 1.5 6.1 Recirculation Loop Elbow 0.768 9.2 Structural Beam (12WF40) 0.768 4.0 Structural Beam (12WF40) 1.5 1.1 With respect to the structural beams, the highest velocity (115 fps) occurs for the 10WF33 beam. For the sake of conservatism, the analyses used a heavier 12WF40 beam with a velocity of 115 fps rather than the actual velocity of 60 fps. 2.1.2 Systems Affected by Line Break Main Steam The MSLs discharge from the vessel at el 310 and at the 90-deg and 270-deg radial locations and descend. In this area there are two 10-in EC lines at the 67.5-deg and 292.5-deg radial locations. These lines are separated by great enough distances so as not to be affected by any break of the MSLs. The MSLs proceed downward through el 295 where they pass by some 1-in instrument piping at the 90-deg radial location and a CRD exhaust line at the 270-deg radial location. Continuing on down from el 295 to el 264, the steam lines pass by containment spray spargers, feedwater lines on each side of the steam lines, relief valve discharge lines, shutdown cooling, reactor recirculation, core spray and rod drive exhaust. All could be ruptured except for the recirculation lines because of their larger size. From this point, the steam lines head toward 180 deg and exit the drywell adjacent to the two 18-in incoming feedwater lines. The required systems for core cooling and safe shutdown in the case of a MSL break are the containment spray, core spray, and NMP Unit 1 UFSAR Section XVI XVI-74 Rev. 25, October 2017 feedwater. All of these systems have the required redundancy or backup as discussed in Section 2.1.3 below. Feedwater The two 18-in feedwater lines enter the primary containment at el 263, adjacent to the two 24-in MSLs. At this point the lines curve around to the 90-deg and 270-deg radial directions. At +/-45 deg on each side of these lines, two 10-in lines proceed inward and then ascend vertically. In this run the feedwater lines pass by some 12-in core spray lines, 14-in shutdown cooling lines, 6-in containment spray lines, and 6-in cleanup system lines. Proceeding upward to el 295, the feedwater lines enter the reactor vessel at the 45-deg, 135-deg, 225-deg, and 315-deg radial locations. During this ascension they pass by the 12-in core spray lines and containment spray spargers, and a 1 1/2-in liquid poison line. The only damage which could occur is to one of the redundant containment sprays and to the liquid poison lines because of their size. As discussed in Section 2.1.3 below, this system has adequate redundancy even in the event that one system is incapacitated. The liquid poison system is a post-LOCA suppression pool pH control and backup reactivity control system only. The liquid poison system is not required to function following a feedwater line break, since the primary reactivity control system (CRD) is not impacted by a feedwater line break. Reactor Recirculation System There are five recirculation pumps, each of which has a suction and discharge line. These lines are at the 0-deg, 42-deg, 73-deg, 114-deg, 144-deg, 186-deg, 216-deg, 258-deg, 288-deg, and 330-deg radial locations between el 225 and 275. These are 28-in and 26-in diameter lines for the suction and discharge, respectively. Other lines in the area of these recirculation lines are: 1. 10- and 12-in ECs (330 deg, 0 deg). 2. 4- and 6-in containment spray (0 deg to 360 deg). NMP Unit 1 UFSAR Section XVI XVI-75 Rev. 25, October 2017 3. 10- and 18-in feedwater (288 deg, 258 deg, 216 deg, 144 deg, 114 deg). 4. 12-in core spray (258 deg, 115 deg). 5. 14-in relief valve discharge (288 deg, 216 deg). 6. 6-in cleanup (42 deg). 7. 24-in main steam (258 deg, 216 deg, 186 deg, 144 deg, 114 deg). 8. 14-in shutdown cooling (330 deg). 9. 4-in, 2-in, 1 1/2-in, and 3/4-in reactor building closed loop cooling (RBCLC) piping supplying recirculation pump seal and motor coolers.* The first four systems (1, 2, 3 and 4) may be required following a break in the recirculation system. All have the required redundancy or backup. Containment Spray Due to the small size of the containment spray lines and the fact that they are not pressurized during normal operation, rupture would not cause damage to any other larger lines. Liquid Poison The liquid poison line, due to its small size, would not impart damage on any other system. Emergency Condensers The original pipe whip analysis considered only the impact on engineered safety features (ESF) required to maintain containment integrity and adequate core cooling. The conclusion reached was that ESF systems have adequate redundancy and separation of piping inside the drywell to ensure the ESF function is maintained. The RBCLC system was not considered an ESF system and was not evaluated. Since the RBCLC system provides a safety-related post-LOCA function, the consequences of a loss of RBCLC on the ability to maintain post-LOCA safe shutdown has been evaluated on a coping basis. The conclusion reached is that adequate provisions exist to either isolate the RBCLC drywell piping and restore the system post-LOCA, or that alternate methods could be implemented to maintain post-LOCA safe shutdown conditions assuming RBCLC is not restored. NMP Unit 1 UFSAR Section XVI XVI-76 Rev. 25, October 2017 The 10-in EC supply lines are located in the area between 270 deg and 90 deg radially at el 306'. In this area there are only some small instrument lines and some 1 1/2-in containment spray headers. The supply lines leave the drywell and the 10-in return lines enter at el 269'. In this there are four lines: 1. 6-in cleanup. 2. 6-in, 8-in, and 12-in containment spray. The only required lines in the event of a break of an EC line are the containment spray lines. These are supply lines to spargers, and damaging one would not render the system inoperable since there are four spargers. Only one sparger could be damaged by the break of any one EC line. Control Rod Drive Discharge Due to the small size of the discharge piping in relation to the other lines, no other damage would be imparted due to a rupture of this line. Core Spray There are two 12-in core spray lines which enter the containment at el 240' at about 45 deg on each side of the 180-deg direction. The one line in the quadrant from 180 deg to 270 deg passes by the following lines and then rises vertically: 1. 10- and 18-in feedwater. 2. 24-in main steam. 3. 14-in relief valve discharge. 4. 6-in containment spray. The other core spray line, in the quadrant from 90 deg to 180 deg, enters the drywell and then runs north to the 0 deg-90 deg quadrant where it rises vertically. This line passes by the following lines: 1. 10- and 18-in feedwater. 2. 14-in relief valve discharge. 3. 6- and 8-in containment spray. NMP Unit 1 UFSAR Section XVI XVI-77 Rev. 25, October 2017 Both lines rise to el 295 where they enter the reactor vessel 180 deg apart. In their rise they pass by 10-in feedwater lines and some small containment spray headers. The only lines which the core spray could damage are the four smaller feedwater lines and the four containment spray lines to spargers. Relief Valve Discharge In each 180-deg radial segment of the drywell that is from 0 deg-180 deg and 180 deg-360 deg, there are three 14-in discharge lines. In the 0 deg-180 deg sector there are the following lines: 1. 6-in cleanup. 2. 10- and 18-in feedwater. 3. 12-in core spray. 4. 6-, 8- and 12-in containment spray. 5. 24-in main steam. In the 180 deg-360 deg segment there are the following lines: 1. 14-in shutdown cooling. 2. 10- and 18-in feedwater. 3. 12-in core spray. 4. 6-, 8-, and 12-in containment spray. 5. 24-in main steam. The lines which could be damaged by the relief valve discharge that are required for core cooling are the core spray, containment spray and feedwater lines. No single relief valve discharge line failure could eliminate redundancy to the point where the safeguards function is inadequate. Shutdown Cooling NMP Unit 1 UFSAR Section XVI XVI-78 Rev. 25, October 2017 In the 270-deg to 0-deg radial location there are 14-in supply and return lines to the shutdown cooling system at el 270. In this area there are the following systems: 1. 10-in feedwater line. 2. 6- and 12-in containment spray. 3. 14-in relief valve discharge. 4. 3-in exhaust from the CRD. The first three systems may be required following a shutdown cooling system line break. However, no line break in this system could result in loss of redundancy in those required systems to the point where the safeguards function is inadequate. Cleanup System In the 0-deg to 90-deg radial location, a 6-in line comes out of one of the recirculation lines at el 263, leaves the drywell, then reenters at el 263 and returns into a feedwater line. The only lines in that area are: 1. 10-in feedwater. 2. 12-in core spray. 3. 10-in EC return line. 4. 14-in safety valve discharge. 5. 6- and 12-in containment spray. Due to the large sizes of these lines, there would be no effect on them because of a rupture of the cleanup system. 2.1.3 Engineered Safeguards Protection The preceding analysis shows that engineered safeguard systems could be damaged as a result of pipe whip. However, in no case is the damage extensive enough to result in loss of core cooling, a safe shutdown capability. As described in Section 2.1.2 above, the feedwater system (high-pressure coolant injection [HPCI]) could be damaged as a NMP Unit 1 UFSAR Section XVI XVI-79 Rev. 25, October 2017 result of a rupture in the main steam, recirculating, core spray, relief valve discharge or the shutdown cooling system lines. In any event, since there are two feedwater lines which are physically separated in the areas of concern, only one could be damaged. In addition, core spray and autodepressurization are the prime sources of core cooling. HPCI is only a backup system. There is no single pipe rupture which could result in loss of feedwater and both core spray systems. There are two independent core spray lines 180 deg apart. These could be damaged by rupture of either the recirculation or relief valve discharge lines. However, because of redundancy and physical separation, only one line could be damaged. HPCI serves as a backup. NOTE: The ability of the redundant core spray line to provide adequacy of core cooling is evaluated consistent with the pipe whip analysis design basis. The containment spray system could be damaged as a result of a rupture in the following systems:

• reactor recirculation
• feedwater
• main steam
• emergency condensers
• core spray
• relief valve discharge
• shutdown cooling There are two containment spray systems, each one consisting of a supply and a set of spargers inside the containment. Both sets of spargers in each containment spray system could be damaged as a result of a single line break due to close proximity of the spargers. This would not result in a loss of containment cooling since the suppression chamber water would still be circulated through the containment spray heat exchangers. Degradation of spray efficiency could occur and would depend on the extent of sparger damage. In any event, some spray efficiency would remain. The EC supply and return lines on both systems could be damaged by a rupture of the main steam or reactor recirculation system lines. However, this system is not required to maintain core cooling. Feedwater, core spray, and autodepressurization NMP Unit 1 UFSAR Section XVI XVI-80 Rev. 25, October 2017 provide the core cooling function in the event of a line rupture within the drywell. The CRD hydraulic system could be damaged by a rupture in the main steam, relief valve discharge or reactor recirculation system. However, should these lines be damaged, the rods would scram on reactor pressure. The liquid poison system, which serves as a backup to the control rod system, is not subject to damage by the same lines. The only line whose rupture could damage the liquid poison system is a line in the feedwater system. However, the liquid poison system is not normally used and is not required to function following a feedwater line break for post-LOCA suppression pool pH control. It is also a backup to the CRD system which is not subject to damage by a ruptured feedwater line. 2.2 Outside Primary Containment All high-energy lines were analyzed to determine the effects of postulated pipe breaks. In all cases, safe shutdown of the reactor can be accomplished and the Station can be maintained in the shutdown condition(30). Table XVI-29 lists the high-energy systems which were analyzed. It was assumed that a line break could occur at any point outside of the primary containment. A pipe break could cause failures of other systems because of pipe whip, jet forces or environmental effects on equipment. Systems which could be affected by various line breaks are listed in Table XVI-30. The design of the Station incorporates a number of features which mitigate the effects of pipe rupture. There are redundant systems and components for each shutdown or accident protection function. The locations of equipment, power supplies, cables and instrumentation for redundant systems are physically separated to preclude common-mode failures. Cables, motors, power boards and other equipment are designed to be operated in the environments expected after a line rupture. 3.0 Building Separation Analysis The building separation was determined in design so that horizontal deflection will not result in the striking of adjacent structures. Figures XVI-46 through XVI-55 indicate NMP Unit 1 UFSAR Section XVI XVI-81 Rev. 25, October 2017 building separation and computed maximum horizontal deflections () of various structures. Bedrock at the site is sound and competent. No permanent relative displacements would occur during any possible earthquake. Class I piping between buildings (e.g., condensate supply to emergency core cooling system [ECCS]) is sufficiently flexible to withstand the relative displacements indicated on Figures XVI-46 through XVI-61. 4.0 Tornado Protection The probability of a tornado occurrence at the Nine Mile Point site is close to zero. The map of tornado probability published by Fawbush shows less than one occurrence in this general area during a 30-yr period(31). The more complete data assembled by Spohn indicate, for the 1916-1961 period, a total of one to three tornadoes in the 1-deg latitude-longitude square surrounding the area(32). Flora provides details on New York State and shows only a single tornado occurrence in the immediate area of the site itself(33). Using work by Thom(34), one could calculate a 1-in-19,000 yr tornado occurrence probability for the Nine Mile Point site area. In the event of a tornado, it would be necessary to maintain the integrity of certain areas of the Station to conduct a safe shutdown. Those areas and the structures surrounding those areas were investigated by determining the pressure and/or wind velocity at which functional failure would occur. The results of this investigation are given in Table XVI-31. Functional failure of a structural member is reached when that member no longer performs in a reliable manner. Working stress design was used for evaluating Station structures. The allowable or working stress was derived from the AISC or ACI applicable Codes*. The stress at functional failure is the working stress times a safety factor of: Structural Steel 1.5 x Fa 0.9 Fcr 0.9 Fy Concrete 1.6 x 0.45 F'c = 0.75 F'c = Fc Reinforcing Steel 1.80 x 0.5 Fy = 0.9 Fy Definitions (all units in psi) Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section XVI XVI-82 Rev. 25, October 2017 Fa = normal design stress (AISC handbook) Fcr = minimum buckling stress of the material Fy = yield stress of the material Fc = normal allowable stress (ACI Code 318) F'c = ultimate concrete strength Internal pressure relief panels in the reactor and turbine buildings are designed to release at 65 psf and 62 psf, respectively. The ratio of relief area to building volume is 1.6 ft2/1000 ft3 for the reactor building, and 0.21 ft2/1000 ft3 for the turbine building. The capability of the metal siding on the buildings to resist tornado-induced missiles is not known. The H. H. Robertson Company, in conjunction with the Gulf General Atomic Corporation, conducted a series of impact tests to determine the capability of their metal panel siding to withstand certain missiles. The results of these tests and the conclusions drawn were not available. Cables for reactor protection and engineered safeguards systems are routed in accordance with IEEE "Single Failure Criteria" via separate paths between the remote equipment and the control room equipment. Two routes are below grade, at el 261; two routes are above grade, at el 261; both sets run north-south at opposite ends of the turbine island. When below-grade cables must be extended to remote equipment terminations above grade (or above el 340 in the reactor building), individual rigid steel conduit is used for protection and isolation of the cables. Graded scales of action, governed by weather conditions, progress from normal Station operation to emergency shutdown. Action is as follows: Condition Green No U.S. Weather Bureau tornado warnings exist within a 60-mi radius of the Station. Condition Yellow U.S. Weather Bureau tornado warning exists within a 60-mi radius of the Station. Condition Red Sustained steady winds above 100 mph. NMP Unit 1 UFSAR Section XVI XVI-83 Rev. 25, October 2017 5.0 Thermally-Induced Overpressurization of Isolated Piping NRC Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," identified the concern that thermally-induced overpressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions, and could also lead to a breach of containment integrity via bypass leakage. Piping system segments considered susceptible to thermally-induced overpressurization are analyzed in accordance with the criteria of the ASME Boiler & Pressure Vessel Code, Section III, Appendix F, 1986 Edition. E. EXHIBITS This section contains inspection reports for, and welding procedures used in, the fabrication of the reactor vessel. Also included are testing specifications for liquid penetrant and ultrasonic weld examinations. Note: Some of the exhibits have been re-created for the purpose of legibility. Reference the Nine Mile Point Unit 1 FSAR or Records Management for original version. Nine Mile Point Unit l UFSAR ' --------......... ------*-----______ . ___ ...., ___________ ---* c:oNetCINI* LUKENS STEEi. COMPANY PHHICAL *enNa LAllOl'tATORY T/9/611 . 11 caATl!ev11..L*. PA. I I . TEST CERTIFICATE ---DAfl PILI NO l!**Me i-OllCMAmiA Ell-T8611-BD -Combu1tlon Iner., Ino._ 10. Chlllttnnooca DlY. Mr. L\1thor 1ow17, '-"* ruroh. Chattaw..osa, 'l'enn. I I I-MELT HO. ra130 MILLORDlll

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• Dl:SCAIPTION /ti II-!Tl. lll?raGB;e YB-P3F6B f2130 1 837 !coo.: tfO* x 135-%" x lllf-3 313 &:imc AD 565 003 31J G-1 1-1 " lJA003:>2-5G or.I ) tol" aa. * .. ,.., [, llod, for nGl't R *If-mo. i, ...... "' * *mtro t11. "* I FbJI, roooo C.G.( oAJ r,.,' le. T ,.. 0 to I *12h*lc1 " I watt*t iau.nch* ln l 5 llt.nt to 00 , ** 'l'C!lt B then tcrnperec 00/125< *** t elcl 1 ... , ... llnch 1 tntmu1
• tumaoe eool ocl to 6oo0 ... "Plat* tum' uhect tn 111 roll ed tt1t1per. * -. . . .. . . -* . . ... NH*lllOtl*ftet.--1_ .. /7,..31 "P..1 //I " UFSAR Revision 16 XVI-76 (EXHIBIT 1 Sh 4 of 4) November 1999 Nine Mile Point Unit 1 UFSAR FABRICATION TEST PltOGRAM FOR RIAaARA MOIWIX
• 213" IWR G!NEIAL ELECTllC COMPANY PURCBASE ORJ>EI RO. 205*09596 'l'ES'1'ED IR ACCORDANCE W1'1'H ATrACBMER'I' C 01 G.E. SPECIFICATION 21Al104, IEV. RO. 1 CERTIFIED BY VENDOR BY DATE.I Ul 7 i;i:S FOR llllHl fll'!9 "llmR A1DlllC ftMll quip. llPJ. .. JlllSL c:AUfOl9IM -----UFSAR Revision 16 CEl.TIFIED AS TESTEJ> IT COMBtJSTIOfi ENGIHEEllING CE CONTIACt NO. 164 , GENERAL ELECTRIC CO. " ANO-MN .109E 1236 -1 43-\ : EPrf
• XVI-77 (EXHIBIT 2 Sh 1 of 10) November 1999 Nine Mile Point Unit 1 UFSAR The test material was removed from a piate, Code G*38. The chemical 0 analysis of the plate is as follows: c 19 Mn 1.21 008 s 018 Si 22 Ni 52 Mo 49 The plate was beat created by CEI at 1500°-1600°F, held at ture four (4) hours, dip quenched in agitated water plus 1225° :t 2s0r, held at temperature four (4) hours. The plate was then given an additional heat treatment of 115<>° +/- 25°F thiny (30) hours furnace cool to 600°r. *
• The teat specimens were removed in accordance with Drawing E-231-586 and were tested in accordance with Ate&chment C to customer's Spec. 21All04, :Rev. l. The following teat data are for Item Code G*38 which is SA*302B plate 75" x 118" x 7 1/8". UFSAR Revision 16 XVI-78 (EXHIBIT 2 Sh 2 of 10) November 1999 0 0 Nine Mile Point Unit 1 UFSAR Qiarpy V Notch Impacts Tut Tut Lateral Code Location Temperature 0r Foot Pounds i Shear Expr.nslon C-7 Plate Surface -60 43.0 10 .033. c-a Plate Surface -60 34.0 10 .028 c-4 Plate Surface -40 44.5 20 .037 C-5 Plate Surface -40 42.5 20 .036 C-6 Plate Surface -40 61.5 30 .050 C-13 Plate Surface -25 52.0 25 .043 C-14 Plate Surface -25 55.0 25 .047 C-15 Plate Surface -25 54.0 25 .047 c-1 Plate Surface +10 68.5 50 .054 C-2 Plate Surface +10 79.0 70 .064 C-3 Plate Surface +10 69.0 50 .054 C-9 Plate Surface +60 84.0 85 .065 c-10 Plate Surface +60 94.0 90. .071 C-11 Plate Surface +uo 98.5 100 .* 073 c-12 Plate Surface +llo 104.0 100 .073 C-22 1/4 Thickness -60 17.0 -o-.013 C-23 1/4 Thlckneu -60 26.0 -o-.017 C-19 1/4 'fhickneH -40 33.0 5 .028 C-20 1/4 Thickness -40 29.0 5 .025 C-21 1/4 Thlckne99 -40 30.0 s .e26 C-29 1/4 Thickness -25 35.5 10 .029 C-30 1/4 Thickness -25 30.5 10 .OZ5 c-16 1/4 Thickness +10 57. 5 40 ,01,7 c-11 1/4 Thickness +10 Jl.5 20 .032 c-1e 1/4 ThlckneH +to 20.0 15 .022 C-23 1/4 Thickness +.to 57 .s 40 .Ot.9 C-24 1/4 Thickness +60 62.0 50 .054 C-25 1/4 Th ickncss +60 69.0 60 .058 C-26 1/4 Thickness +110 89.0 95 .Q7l C-27 1/4 Thickness +110 104.0 100 .OBl c-34 Center Thickness -40 20.0 5 .016 C-35 Center Thickness -40 10.5 5 .013 UFSAR Revision 16 XVI-79 (EXHIBIT 2 Sh 3 of 10) November 1999 Nine Mile Point Unit 1 UFSAR t '*": :**
• J1t 1n 1m. 1N,:11 ' f
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• I '. I ' ** * .. . , I .. . . ** . . .. . . 1!,'1 !!ii :1'ti1*:11:11:il::,.!!1'1!I 1111 ** I !111111 ' I !!! I !' ! 111.1 II I "I Iii'. '* . *U 11; 1 ** I :* .* _:* ! ' '\ !.. *l: 111. 1,11 II : 111 If: I t *, * * *:t. I ' * 'I Ii'; '11111 10 * . , I ' :1 ' I! . . ** ; I . II ,. I i . 1! 'f  :* ... ,u. 11 im 1!!! !!I: !:11 ,1'*. . ! :t : *!' ,rjl !I!! 1ili !1'1 !!I! *!Ii ;l!i ;tirl!k ,, I! ii!, 11,,1 !!*1 ::! 'ii; :11*1 !:.; .,.,: .,*11 II" *! , ..... , ... ,1,; ,,, I 0 I ** I* !,;* 1f t *1t llt *1' , J 1:r1 1111 tltl Hj titl :1u uI! m. iJ -i" i!:: '* m** , 1*11111* *111 *111.. . ... ... .. .,, *, *;* *:!'. 11 ...... < :,!.! !' ! i : *lL ;n1 hi* 011: .ii *=' ,, . ;", * : , . I , -. , , . , I I ttl .. * , .. *** --. I ... L:..:.J..i...l::.:k:...I"* ..... l*t:'!ll:'l'!lll'l"llllll;*l'I 111:"l1'h!** Tli*!ij\'llli:J:l'T*'*!IUlllUll!.!0 i I' 1fl: 11!1 :l!i *l.1111: *1.1 '!.! 1:1! li!' .*ii i!! lH !i ':'* j\i' ..:..J!p o .,.,,,0 so ,;i..,_ ° ,;lee IO UFSAR Revision 16 XVI-80 (EXHIBIT 2 Sh 4 of 10) November 1999 Nine Mile Point Unit 1 UFSAR CIJarpy V Notch Impacts (Continued) Teat C-36 c-45 C-37 C-38 c-:n C-32 C-33 C-39 C-40 c-41 C-42 C-43 C-44 C-49 C-50 C-51 . C-58 C-59 C-60 C-52 C-53 C-46 C-47 C-48 C-54 C-55 C-56 C-57 Location Center Thickness Center Thickness Center Thickness Center Thickness Center Thickness Center Thickness Center Thickness Center Thickness Center Thickness Center Thickness Center Thlckness Center Thickness Center Thlcknesa 3/4 Thickne&1 3/4 Thickness 3/4 Thickness 3/4 Thickness J/4 Thickness J/4 Thickness 3/4 Thickness J/4 Thickneaa J/4 ThickneH 3/4 Thickness J/4 Thickness J/4 Thickneaa 3/4 Thickne911 J/4 Thickness 3/4 ThickneH Test 0 Temperature F 25 20 +10 +10 +10 +60 +60 +110 +110 +160 +160 40 25 25 20 +10 +10 +10 +60 +60 +110 +110 Foot Pounds 16.0 17.0 26.0 34.5 33.0 44.0 37.0 44.0 49.0 77. 5 85.0 92.0 80.0 28.0 .18.5 13.0 29.0 Jl.O 40.0 34.0 34.5 50.0 53. 5 49.5 76.0 52.5 92.0 100.5 l Shear 5 5 10 10 15 20 15 35 40 90 95 100 100 5 5 5 10 10 20 10 10 25 25 25 70 50 100 100 Lateral .017 .016 .024 .023 .030 .038 .033 .041 .042 .053 .C69 .079 .066 .025 .Oi9 .017 .023 .028 .034 .029 .030 .Qlt4 .01, 5 .042 .065
• 01, 7 .069 .082 The following graphs (Dl, 12, 03 and 14) are results frQm Charpy V Notch lmpact specimen-of base material for determining the JO ft-lb. transition temperature at the O.D., 1/4 T, 1/2 T, and T thlcknesa level1 of the plate material. UFSAR Revision 16 XVI-81 (EXHIBIT 2 Sh 5 of 10) November 1999 Nine Mile Point Unit 1 UFSAR I*" :"' .*u *' .... 1 .. to: t .f,f t;? .. n ' ******* 0** * . * .......... . ** tliflfl <llo ***
• t Cft ** , .... ,.. .... :J UFSAR Revision 16 XVI-82 (EXHIBIT 2 Sh 6 of 10) November 1999 Nine Mile Point Unit 1 UFSAR .. :mr INCH *1v1*r1. * ***** ce. ""t!'1J' .. ** n****** ..o .. 1.11*J *i lrl****I**
• l,;!!!1:.!.:.Je UFSAR Revision 16 XVI-83 (EXHIBIT 2 Sh 7 of 10) November 1999 p: II( 0 Ill ...... rz. ::::> IH 0 ...... (X) +I *.-! .c: c Ill ::::> N +I c E-4 *.-! H 0 r:Q c.. H :i:: Q) O* >< .-1 ** rz:i ** *.-! -* ::i: OJ ** qo Q) I: * (X) c . I u * *.-! ! : H :z !. : > >< os J ;j i :! ; ** ... .. ._ -.

Nine Mile Point Unit 1 UFSAR .SOS Tensile Data from Piece "F" Test: Test Yield Strength Ultimate Tensile Elongation Reductlon Code Location Temperature 0r KSI Strength KSI in 2" X of Aren '7. T-1 Pl Surface RT 64.5 85.5 26.0 68.0 T-2 Pl Surface RT 64.5 85.5 27.5 68.5 T-3 Pl Surface RT 64.5 85.5 27.5 68.5 T-4 Pl Surface 550 57.0 83.0 22.0 62.0 T-5 Pl Surface 550 58. 5 83.5 21.5 63.5 T-6 Pl Surface 550 58.5 84.0 23.5 63.5 T-7 Pl Surface 650 55.5 82.0 21.0 68.5 T-8 Pl Surface 650 55.5 82.5 28.0 67.5 T-9 Pl Surface 650 56.0 83.5 25.0 66.5 T-10 1/4 T RT 63.0 85.0 26.0 61.0 T-11 1/4 T RT 62.5 83.0 27 .5 66.0 . T-12 1/4 T RT 63.0 84.0 26.0 65.0 T-13 "../4 T 550 56.5 et.; 21.0 59.0 T-14 1/4 T 550 56. 5 81.5 22.0 61.5 T-15 1/4 T 550 56.0 80.5 21.0 61.0 T-16 1/4 T 650 54.5 82.5 24.0 55.0 . T-17 1/4 T 650 54. 5 79.5 n.o 63.5 T-18 1/4 T 650 54.0 80.0 25.5 59.0 T-19 Center T RT 64.5 88.5 24.0 61. 5 T-20 Center T RT 64.5 87.5 24.0 63.5 T-21 Center T RT 65.0 87.5 23.5 61. 5 T-22 Center T 550 58.5 85.5 11.0 23.5 T-23 Center T 550 57 .5 85.0 20.0 50.5 T-21* Center T 550 58.0 81.0 10.0 32.5 T-25 Center T . J50 56. 5 82.0 19.0 42.0 T-26 Center T 650 56.0 78.0 9.0 20.5 T-27 Centl!r T 650 56. 5 82.0 17 .o 37.0 T-28 3/4 T RT 62.5 84.5 27.0 66.0 T-29 3/4 T RT 62.5 84.0 27.0 66.0 T-30 3/4 T RT 62.5 84.0 27 .o 6'j. 5 T-31 3/4 T 550 54. 5 80.0 22.0 60.0 T-32 3/4 T 550 56.0 81.5 22.0 60.0 UFSAR Revision 16 XVI-85 (EXHIBIT 2 Sh 9 of 10) November 1999 Nine Mile Point Unit 1 UFSAR

• 505 Piece "F" (Continued) Test Test Yield $trength Ultimate Tensile Elongation Reduction Codq, Location I_e111perature 0£ KSI _ StrelU!.th KSI___ of_Area 7. T-33 3/4 T 550 56.0 81.5 23.0 62.0 T-34 3/4 T 650 56.0 79.0 25.0 61.0 T-35 3/4 T 650 55.0 81.0 25.0 62.0 T-36 J/4 T 650 5J.5 79.0 25.0 60.0 Prepared by: ,{j ,/ ;;;::,; UFSAR Revision 16 XVI-86 (EXHIBIT 2 Sh 10 of 10) November 1999 Nine Mile Point Unit 1 UFSAR :c: . Mr .* R. ? . I.=renu C. *=t. **M:; B ._ Cc-ilJter L..LURGICAL. F-ESf AF.CH ANO CE\'Q.DPME::T O:FT. *-VESSEL WELD iFST .. Mr* 1
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• General Electric. A!'E:D CQS ____ ...... ....,..-...--..;;;;,;;,;;;__.,,;;,,;;;......., ...... -----*CONTRA.CT NO. !9665 .j,1.A. TERIA.L ___ .SA.._-_3_02_-_B_M_od_. -----'."""1-_3_* __ .JOB tJC. P-31983 ___ i-_2_3_2-_0_1_2 _______ _ SEAM NC. __ ._2-_0_7_2_A_-_r __ _ DETAIL WELD PROCEDURE. NO. _._._.wa ..... * ... s,_-... o .... .. . . . TK 8-1/2" FlL.L:.ER UETAL 0(TYPE, HT. AND SIZE) . 3il6.RACo 3. m.,No. W-5%1(, .l/16 NI-200,
• m. No. N-77S3A, Back Weld 3/16 & 1/4 i'll-i'i!Jt *.FLUX (TYPE AND LOT) OR GAS.. ..Liilde 1092-65-200 Lot 361-7 * . . . . .. . .. ** . . : . . : .. **:rEMP:. *.* 1150* F-: . . . : . . . : . ,,... . -: ")""* . --. .... :. :* ** ... .. KRS._...,co ______ _ . . *.. .. * ; ... , ... ** .. -. .. :r -. .. *. *-* . ALL WELD0MET.A.L AND .. "OR TRANSVERSE JOINT REDUCED.SECTION TENSILE TESTS . 'AHA
• IA.TUT! ID. IN . . LOAD W TTPE DIM!tWCllCS . . PEQ.'.:!H n7nC TKlCIOP..SS ;' *.i;ransverst : .... " ---*-.; "'---*1-.. .. __. tflliL. -*. *\.. .... Ull:ATIOH . I TP'. .****: ... ** -....* -: . * * !E).'D TESTS *. . *. . *UJULTS .* * . -.. . Side Bend YALU!S .* .. _ . ..... ..... : ., . * . .* * .. .-. :' .. . . .. .. *. *. *
• WELD DEFT. OR Mu clear Shop *
• NON-DESTRUCTIVE T:STS
• WELDERS' SYMoOL.S TU. YC , .. AG'!; AOG * . . * ." ... .. ." *. :-u-.-. T_Br_NO_:-_*_*_E_-_7..,1_8.6 _____ ,..... ___ _ . : :."*:* *.* . Met. Rese&:-ch & --------------------------------------. ** *91tl'7 r:. at:tr.ta!a .. l!li1 ,._ ....... ee .. , " lut *e!tla ww:. -lu4 *.l taa:1.! ho w!!!i ,._ * . l!T M & P 2.4.?:4.Ca) PT-.----------------* .,.._,, eF Iii. AS!!! Ciiia. RT.
• M & P 2.4.1.3.(a) UT-----------------*-r-tm UFSAR Revision 16 XVI-87 (EXHIBIT 3 Sh 1 of 2) November 1999 Nine Mile Point Unit 1 UFSAR *Transverse to We!c! !enslles 0 Dimensions t1ltim:te e:'!= Test Cede Thickness &a. 1.oad , !.b s
• load, PSI Chare:t"!:-o.f Fai!Uf4? XHBl 1.004 2."023 2:031 170,000 83.7 Ductile in Viele! *12 1.003 1.951 1.957 1S3,250 83.( * "13. 1.000 1.925 1.925 161,750 H.O *
• B4 i.010 1.858 1.876 83.2 * .Al l.000 2.cos 2.005 liO,SCU --* -0 ... 11.1 A2"* .997 1.865 l.8S9 155,000. 83 .4 ** .. A3
• 998 2.040 2.03S . 171,000 84.0 *
• A4. .* !97 1.820 1.814 151,500 83.5 *
• Drop Weiqht Tests Weld Temti. *1 Results NPT -*1 -10 No Failure Less .Than -so* F -30 *. -so * -so
• Heat Affected Zone 0 -10 No Failure .. Less Than -so* F -so * -so * -so
• 0 UFSAR Revision 16 XVI-88 (EXHIBIT 3 Sh 2 of 2) November 1999 llft Nine Mile Point Unit 1 UFSAR WELDING PROCEDURE SPECIFICATIOIJ COMBUSTION ENGINEERING, INC. .tppentt;z I 13. SPEC'N. NO. DATE SEPT>>m!R l"6, 1958 PROCEDURE FOR WELDING TS;._..;.LOW __ ALLOY_;;;;;___s_1_*e_:e_:r_s ____________ _ WELDING PROCESS: THE WELDING PROCESS SHALL BE MANUAL SHIELDED METAL ARC USING CONU:mlOOAL MAl.flJ& SHIELDED METAL ARC TH!CICNESS: THE THICIOCESS COVERED BY THIS SPECIFICATION SHALL BE .-N,_O_RES=;.;TRI=.-.CTI=-0'-"NS;;;;.-______ _ BA$E *!£RIAL; f.:. p THE TYPE CF *TERIALSHALL BE _PIPE, PLATE, FORGINGS OR WELD CLADDING CONFORMING TO SPECIFICATIONS SEE ASME SECTION IX TABLE Qll.l tJ.
• AND To ASME Ca>E, SECT. IX. P*NCL 3 v N 2:J lll":n OR TO THE FOLLOWING ANALYSIS; lJn Ho OR Mo -______ WELDING *TERIAL:
• THE TYPE OF FILLER METAL SHALL BE _ ]$ NICKEL CONFORMING TO SPECIFICATION _ ASME -SA3lo:-i80i5=16Q AND FOR METAL ARC, TO ASME CODE, SECT IX, F*NO. 4 AND A*NO. -*--3.._. __ _ OR TO THE FOLLCJ#ING ANALYSIS; ----------------------------FOR OXYACETYLENE WELDING THE FILLER a.ETAL SHALL BE (SI) (AL) ( ) KILLED. FOR SUBMERGED ARC WELDING THE FLUX SHALL BE ---------------------FOR INERT *GAS ARC WELDING THE SHIELDING GAS SHALL BE WELDING PROCEDURE : THE WELDING POSITION SHALL BE _NOT FOR *CHINE WELDING THE SHALL BE PASS METHOD SHALL BE USED, .At<<> THE Nt.NBER OF ARCS BACKING STRIP OR RING REQUIREMENTS SHALL BE OTHER REQUIREMENTS ------AS DETAILED ON PRODUCTIOirlfil.WliGS ___ _ PREHEAT: THE PREHEAT AND INTER*PASS TEMPERATURE SHALL BE _J.§Q2MINIMUM -SEE "OTHm POSTt£AT: THE POSTHEAT TREATMENT SHALL BE FINAL TREATMENT: 11$0°F + 2$°F I HOLD ONE HODR INCH_.QL'llfICKNESS. OOOL TO 600°F, INTERSTAGE: 1100° F -1179' F HOLD J.5 MINUTES TIME. NON* ESSENTIAL BASE *TERIAL PREPARATION: THE BASE MATERIALJaNT PREPAR.ATION*YBEBY MACHlNING OR GAS CUTTING ____ AND g_RIPP.!N..Q..A?:!g mrnrnmo. ELECTRICAL CHARACTERISTICS: THE WELDING CURRENT SHALL BE DC -RP OR AC MAY BE USED FOR E-801:6 ELECTRODES WELDING TECHNIQUE: NORMAL APPEARANCE OF LAYERS: THE WELDING TECHNIQUE SHALL BE SUCH THAT THERE SHALL BE PRACTICALLY NO UNDERCUTTING OF THE BASE METAL. CLEANING: ALL SLAG AND FLUX SHALL BE REMOVED BEFORE EACH BEAD. DEFECTS: CRACKS OR HOLES SHALL Bl! EXCAVATED BEFORE EACH BEAD. TREATMENT OF ROOT AREA: BY PRODUCTION,--=:IIIU:=:wIN=.::GS::.*:.-------------------*------------------------------------------INSPECTIOI REQ!JIREMENTS: THE SURFACE FINISH AND INSPECTION REQUIREMENTS SHALL BE AS REQUIRED BY JOB SPECIFICATI CHS AND DRAWINGS OR AS FOLLOIS, OTHER REQUIREMENTS: ...... ovm 2" ....J!Qlfil:§.. ov:m l" IN THICKNESS H::JLD THE PREHEAT TJ!l'!PERATURE UNTIL POSTHEAT IS STARTED. *-----------*----COMBUSTION INC. BY /:LLY.L_Y1lrs . j UFSAR Revision 16 XVI-89 (EXHIBIT 4 Sh 1 of 2) November 1999 31Sr Nine Mile Point Unit 1 UFSAR WELDING PROCEDURE COMBUSTION ENGINEERING, INC. .&ppend1z I 13. SPEC'N. NO. PROCEDURE FOR WELDING
• 80,000 TS LOW AlJJJ'! STE!LS DATE SEPTEMBER ""J.'6,.-;,;...,,.,19...,S-8-. WELDING PROCESS1 THE WELDING PROCESS SHALL IE MANUAL USING CONVENTICIUL MANUA&. SHIELDED METAL ARC EQUIPME!!t,,,.,,.""""......,-------TH!C!CNESS I THE THICKNESS COVERED IY THIS SPECIFICATDI SK\LL IE ... N...,O_RES ___ m __ c_n__,o ... NS..._ _____ _ WE MllJRIAL I THE TYPE OF M4TERIAL SHALL BE PIPE, PLATE, FORGINGS OR..J!ELD CLADDING ,.. .. s. P. CONFORMING TO SPECIFICATIONS SEE ASME SECTION IX TABLE Qll.l AND TO ASME CCDE, SECT. IX. P*NOS. .J OR TO THE FOLLOlflNG ANALYSIS: Qr Mo (CASE -1236 -SA-336) JfJJV 231159. WELDING MATERIAL:
• THE TYPE OF FILLER METAL SK\LL BE U NICKEL CONFORMING TO SPECIFICATION ASME -SA3lo:EE!Oi5-l6G ANO FOR METAL ARC. TO ASME CODE, SECT IX, F *NO. ---.:s------AND A*NO. --'"-----OR TO THE FOLLCJrlNG ANALYSIS: FOR OICYACETYLENE WELDING THE FILLER METAL SHALL BE (SI l (AL) ( ) KILLED. FOR SUBliERGED ARC WELDING THE FLUX SHALL BE FOR INERT *GAS ARC WELDING THE SHIELDING GAS SK\LL BE -----------:--------WELDING PROCEDURE: THE WELDING POSITION SHALL BE _..,H.::.OT:...:.RES=.::TRI=C._.TED=-----------------FOR MACHINE WELDING THE PASS METHOD SHALL BE USED, AN> THE Nl.leER Of ARCS SHALL BE BACKING STRIP OR RING REQUIREMENTS SHALL BE AS DETAILED ON PRODUCTION IRAWlNClS OTHER REQUIREMENTS ---------------------------PREHEAT: THE PREHEAT AND INTER* PASS TEMPERATURE SHALL BE MINIMUM -SEE *OTHm REQ1iiii:p!NTS* POSTHEAT: THE POSTHEAT TREATMENT SHALL BE FINAL TREATMENT: 11 0°F + 2 °F1 HOLD ON! HOUR INCH OF THICKNESS. FURNACE CX>OL TO 600°F. INTERSTAGE: llOO"F -ll7S°F HOLD l5 MINUTES TlME. NON* ESSENTIAL VARIABLES BASE MATERIAL PREPARATION: THE BASE MATERIAL JCINT PREPARATION M4Y BE BY MACHINING OR GAS CUTTINO _____ AND 9._RIPPlNJL!@ ... ____ _ ELECTRICAL CHARACTERISTICS: THE WELDING CURRENT SHALL BE DC -RP OR AC MAY BE USED FOR E-8016 ELECTRODES WELDl< BEFORE EAOt BEAD. DEFECTS: CRACKS OR HOLES SHALL BE! EXCAVATED BEFORE EACH BEAD. TREATMENT OF ROOT AREA: AS REQUIRED BY PRODUCTION mAWINGS. INSPECTION REQUIREMENTS: THE SURFACE FINISH AND INSPECTION REQUIREMENTS SHALL BE AS REQUIRED BY JOB SPECIFICATI CJG AND DRAWINGS OR AS FOLLOIS. DrHER REQUIREMENTS: JOINTS OVER 2* IN mrmoo:.ss AND LONQllUJ2lliAI,,. JOINTS ovm i* IN THICKNESS HOLD THE UNTIL POSTHEAT IS STARTED. COMBUSTION ENGINEERING, INC. BY U' 7S. -6 ":F! I UFSAR Revision 16 XVI-90 (EXHIBIT 4 Sh 2 of 2) November 1999 0 0 Nine Mile Point Unit 1 UFSAR WELDING PROCEDURE SPECIFICATION COMBUSTION ENGINEERING, INC. C-0-P-Y SCOPE I PRCX:EDURE FOR WELDING 80,,000 TS UM ALLOY STEELS Appendix I 3l. SPEC'N. NO ... DATE SEPmiBER 16, 19i!: PROCESS: THE WE.LDJKG ?-!A CHINE -______ _ USING TH!CKNE$S: THE THICKNESS COVERED BY THIS SPECIFICATIOC SHALL BE NO RESTRICTION BASE MATERIAL; THE TYPE OF MA. TERIAL SHALL BE CONFORMING TO SPECIFICATIONS PIPE, PLATE, FORGrnGs OR WELD CLADmNG SEE ASME SECTIO!l IX TABLE Qll.l 3 Al<<> TO ASME COOE, SECT. IX, P*NOS. CR TO THE FOLLOWING ANALYSIS: Mn Mo OR N,..i._-_c __ r-.... M .... ... s ... E .... l,...2..,3 .... 6_-___ s __ A __ -.. 33._.6,..). _____ _ WELDING MATERIAL; THE TYPE OF FILLER METAL SHALL BE ...... .l. ... 2%..._.Mn......., * .___1/ .... 2....,% ....... Mo,.__,_.( ... RA .... c ... o..__N .... o __
• _3..__0R..._.E...,Q...,UIV=--A--LE=NT-)..._ __ _ NONE APPLICABLE CONFORMING TO SPECIFICATION ANDFORMETALARC:. TOASMECODE, SECT IX, F*NO. ANDA*NO. OR TO THE FOLLOWING ANALYSIS; FOR OXYACETYLENE WELDING THE FILLER METAL SHALL BE (SI) (Al) ( ) KILLED. FOR SUBtE RGED ARC WELDING THE FLUX SHALL BE ARCOS B5 FOR INERT *GAS ARC WELDING TME SHIELDING GAS SHALL BE WELDING PRCX:EDURE: THE WELDING POSITION SHALL BE FLAT,__ ____ _ FOR MACHINE WELDING TME MULTIPµ:=-----PASS METMOD SHALL BE USED, At<<> THE NUMBER OF ARCS SHALL BE .. .,..ONE.....__ ---------*--------BACKING STRIP OR RING REQUIREMENTS SHALL BE DRAWI._N-!A2...___. OTHER REQUIREMENTS ----PREHEAT: THE PREHEAT AND INTER* PASS TEMPERATURE SHALL BE P05Tt£AT: THE POSTHEAT TREATMENT SHALL BE .. rnMJ. ,IREATI-::NT: uso* .:!:. .?5..F HOLD tml'!MUM ONE .. -=.;.;..;;;;-...H""OUR PER mCH OF THICK::ESs. FUR1':ACE COOL TO 600°r. INTERSTAGE: 1100 -111s*r HOLD 1ITm11!TEs rmrnfu.i Til:E. NON-ESSENTIAL VARIABLES BASE MATERIAL PREPARATION; THE BASE MATERIAL JaNT PREPARATION MAY BE BY MACHINIHG, __ __ c_D_111NG AND n1q_ Al!JL'"..zRillQ!IiG. ELECTRICAL CHARACTERISTICS: THE WELDING CURRENT SHALL BE ALTERNATIUG WELDt<<a TEOINIQUE: )!ULTlPLE PASS -....... l._78.-.".....-T .... HI""'C""K..._ _________ _ APPEARANCE OF LAYERS: .THE WELDING TECHNIQUE SHALL BE SUCH THAT THERE SHALL BE PRACTICALLY NO UNDERCUTTING OF THE BASE METAL. CLEANllG: ALL SLAG AND FLUX SHALL BE REMOVED BEFORE EACH BEAD, DEFECTS; CRACKS OR HOLES SHALL Bl! EXCAVATED BEFORE EACH BEAD, TREATMENT OF ROOT AREA: AS REQurRED BY PRODUCTION INSPECTION REQUIREMENTS: THE SURFACE FINISH AND INSPECTION REQUIREMENTS SHALL BE AS REQUIRED BY JCB SPECIFICATI tJG AND DRAWINGS CR AS FOLLOIVS, --OTHER REQUIREMENTS: J'CR 2'ilfNTHICKNrSS AND LONGITUDINAL JdlliTS OVER 1" IN THICKNESS HOLD THE PREHEAT TEMPERATURE UNTIL POST HEAT IS STARTED. --COMBUSTION ENGINE.ERING, INC. POllllll* .... **Y. BY w. B. BUNN UFSAR Revision 16 XVI-91 (EXHIBIT 5 Sh 1 of 1) November 1999 Nine Mile Point Unit 1 UFSAR WELDING PROCEDURE SPECIFIC>. TICN COMBUSTION ENGINEERING, INC. Appendix I 32. SPEC'N. DATE AmusT 3, 19$9 PROCS>URE FOR WELDING 4l§TEHITXC HELD CLADDING 80,000 TS UM JUDJ STIE§ WELDING PROCESS1 THE WELDING PROCESS SHALL BE HAHUAL SHIEUlED METAL JRC THICKNESS 1 THE THICKNESS COVERED BY THIS SPECIFICATICJt SHALL BE ..:.N;;:OT::....:RE:l=:.:TRI=C:.:-T::ED::;... ______ _ BASE MATERIAL: THE TYPE OF M4TERIAL SHALL BE
• PIP.!, PLATE, lORGINQS OR WELD CLADDIN CONFORMING TO SPECIFICATICH SEE ASHE SECTION IX, TABLE Qll*l AND TO ASME Ca>E, SECT. IX. P*NOS. 0R TO THE FOLLOWING ANALYSIS 1 Mn-Mo OR Ni-Cr-Mo (CASE 1236 -SA336) WELDING M4TERIAL1 THE TYPE OF FILLER METAL SHALL BE
• FIRST LAYER TYPE 309; SUBSEQllmT LAYERS TYPE JC8 _ CONFORMING TO SPECIFICATION _ASKE SA-298 E-.309-lS AND E-JOS:lS ANDFORMETALARC:, TOASMECOOE,..,SECTlX, f*t(O. _-2,, .*7
• CR TO THE FOLLOWING ANALYSIS a lo Cr 6 Hi j.CARBON MAI. AT lfli* CLADDlHO THICKNESS) FOR OXYACETYLENE WELDING THE PILLER METAL SHALL BE (SI) (ALJ ( ) KILLED, l'CR SUBaERGED ARC WELDING THE FWX SHALL B! l'OR INERT *GAS ARC WELDING THE SHIELDDG GAS SHALL BE WELDING PROCEDURE 1 FLAT I HORIZONTAL OR" VERTICAL THE WELDING POSmON SHALL BE FOR M4CHINE WELDING THE """--------PASS M!THCDSHALL BE USED, AND THE N\MER Of ARCS SHALL BE -------------................... -------------BACKING STRIP OR RING REQUIREMENTS SHALL IS! NOT REQUIRED OTHER REQUIREMENTS ----,----------------------l'ltEH!AT1 THE PREHEAT AND INTER*PASS TEMPERATURE SHALL BE i.S.Q* :tQ 4QQ*r -JlOI.1> Jmm PQST 1fft4l' l'CISTHEAT I THE POSTHEAT TREATMENT SHALL 8! FINAL TREATMENT: mo*
• 22*?; HOLD MINIKtll ONE soua .m Ui(ili. er T!!Ttj1Qif'S!5, FtwCJ CX>OL TO 6oC7* nam'UGI: uoct1 -J.J,15*; HOLD l5 MINU'1'ESMINIMt TlME. NON*ESSENTIAL VARIABLES BASE M4TERIAL PREPARATION 1 THI! BASE M4TERIAL JCINT PRl!PARATION M4Y BE BY M&r:HIKT'NG OB GRINDING ____ m.......,,..pRli\lmllil&..A CIJ::&N SUBFACE l!LECTRlc;AL CHARACJJ!USTIC:S 1 THE WELDING CURRENT SHALL Bl! AND VOLTAGE OH ----=THE=-=LOW::.:, SIDE OF THE RANGE. WELDIG TECHNIQUE I usE"STiUif§ER,_B ... EAD ___ TEC _____ HNI __ Q""'UE--. _______________ _ APPEARANCE OF LAYERS 1 THE WELDING TECHNIQUE SHALL BE SUCH THAT THERE SHALL BE PRACTICALLY NO UNDERCUTTING OF THE BASE METAL. CLEANING: ALL SLAG AND FLUX SHALL BE REMOVED BEFORE EACH BEAD. DEFECTS: CRACKS OR HOLES SHALL Bl! EXCAVATED BEFORE EACH BEAD. TREATMENT OF ROOT AREA: NONE REQm:IlED=..------------------INSPECTION REQUIREMENTS: THE SURFACE FINISH AND INSPECTION REQUIREMENTS SHALL BE AS REQUIRED BY JCB SPECIFICATI OHS, DRAWINGS AND AS FOLIDW!L llISDAIJ:Y INSPECT EACH MYER FOB St!RFACE DEFECTS. TEST COMPLETED WELD AFTER FINISH MACHININ:::.:::.:G::..*:.-.__,-----OTHER REQUIREMENTS: AJ;!!S_l!}.Y BE MADE TO THE SECO!!D OR I.Arns, 'W]!!!_! MllIMOM INTERPASS OF 1009°F, PROVIDI!iG THERE IS A MINIMlM OF ]J81i" Ovmu.AI BELal REPAIR __ 10 REPAIRS ARE TO BE MADE OH THE CLADDING AFTER FINAL STRESS RELIEVING WHEm: THE AREA IS OVER 2 SQ. INCH&S AHD THE OVERLAY BELOW' THE REPAIR IS UHIIER 1/8* THICXNESS. *: COMBUSTION ENGINEERING, INC. UFSAR Revision 16 XVI-92 (EXHIBIT 6 Sh 1 of 1) November 1999 0 0 Nine Mile Point Unit 1 UFSAR Appendix I 33
• WELDING PROCEDURE SPECIFICATION COMBUSTION ENGINEERING, INC. SPEC'N. HO. DA TE PROCEDURE FOR WELDING WELD __ WELDING PROCESS: THE WELDING PROCESS SHALL BE USING ARC ------THICKNESS; THE THICKNESS COVERED BY THIS SPECIFICATl(Jj SHALL BE NOT RF.:STRICTED BA,\! MATERIAL: THE TYPE OF MATERIAL SHALL BE __ PIPE, PLATE, FORGmGs OR WELD.-::-CLA::--D_D_m_a _______ _ CDNFORMINGTOSPECIFICATIDNS SEE ASME SECTION ll, TABLE Qll.l AND TO ASME CODE, SECT. IX. P*NCL 3 -------OR TO THE FOLLOWING ANALYSIS: (.Q.ASE _______ _ WELDING MATERIAL: THE TYPE OF FILLERMETALSHALLBE TYPE ,309; TYPE 308. CONFORMING TOSPECIFICATIDN AND ER.308. _______ _ AND FOR METAL ARC. TO ASME CODE, SECT IX, F *NO. _ _]_ _______ AND A*NO. ___ l.__ __ -OR TO THE FOLLC711tlNG ANALYS1S1lfL,P..r_1,J!i_tg_.AJY!.Qti..Q.*_Q6! FOR OXYACETYLENE WELDING THE FILLER METAL SHALL BE (SI) (AL)_ ( ) KILLED. FOR SUBMERGED ARC WELDING THE FLUX SHALL IE FOR INERT *GAS ARC WELDING THE SHIELDING GAS SHALL BE WELDING PROCEDURE 1 THE WELDING PCSmDN SHALL BE _FIJ.:J' _________ _ FOR MACHINE WELDING THE ......m:tm.JL PASS METHOD SHALL BE USED, AND THE NLMIER OF ARCS SHALL BE __ .QD. * .(SERIES.l ____ *-**-----_.:*-*---------* BACKING STRIP OR RING REQUIREMENTS SHALL BE ---** OTHER REQUIREMENTS **------*--------*--*----------llREH;-;:-T..'°EmHEAT AND INTE;:;ASS _ _4.QQ*i .. P05THEAT1 THE POSTHEAT TREATMENT SHALL BE ll2Q:t.!...§.*F_, HOLD otIE HOI!i PER IliCli...OI . ...6.0.0.:I. .. IN'.mS.um:.:......ll.Ql2.':-U75*L HOLD 15 MINl!l'ES MINIMUM TIME. NDN*ISSENTIAL. VARIABLES BASE MATERIAL PRU'ARATION 1 THE IASE MATERIAL JCINT PREPARATION MAY B! BY lf.&.CHTNING QR WBlUlWfG ... .te ___________ , EL.EC!llllCAL 01ARACTERISTIQ 1 THE WELDING CURRENT SHALL BE _,A_C __ M'Bm m--wELDING TECHNIQUE 1 -------------**------* _:sJ..3 * .... ___________ _ APPEARANCE .Of LAYERS 1 THE WELDING TECHNIQUE SHALL BE SU04 THAT THERE SHALL BE PRACTICAL.LY NO UNDERCUTTING OF THE BASE METAL. CLEANING: ALL SI.AG AND FLUX SHALL BE REMOVED BEFORE EACH BEAD. DEFECTS: CRACKS Cit HOLES SHALL Bl! EXCAVATED BEFORE EACH BEAD. TREATMENT OF ROOT AREA: --------------------------------------------INSPECTION REQUIREMENTS: THE SURFACE FINISH AND INSPECTION REQUIREMENTS SHALL BE AS REQUIRED BY JCB SPECIFICATIONS 1 DRAWINGS AND AS FOUDWS.. ___ Y.IS.11.W.I. INSPECT .EA.CH. I.A.YER FOR SURlJ.C! m:EE:C.l'S....._n.uE!iiAN.!_.T.EST_..COHP.LE'.CEIUIELD...Af.lEB._.F.l.NISlUf.lC!mlIN.G.a....----fWfTH-OTHER REQUIREMENTS: TO __ LA_!!:RS ----A MAXDUJM !NTERPASS OF 100-F PROVIDmG THERE IS A MINIMUM OF 110* OvmLAY m:ur*JNI) Tl WELDMENT INC. UFSAR Revision 16 XVI-93 (EXHIBIT 7 Sh 1 of 1) November 1999 NMP Unit 1 UFSAR UFSAR Revision 16 XVI-94 (EXHIBIT 8 Sh 1 of 4) November 1999 COMBUSTION ENGINEERING, INC. CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.3.1(a) DATE: April 27, 1960 SHEET: 1 of 4 PROCESS SPECIFICATION FOR LIQUID PENETRANT TESTING 1.0Scope: 1.1This Process Specification provides for the method and standard of acceptance for liquid penetrant testing of nuclear components. 1.2The penetrant test method is used for detecting the presence of discontinuities in ferrous and nonferrous materials. Discontinuities not open to the surface will not appear since penetration into an open defect is necessary before this method will work. For this reason its use is generally limited to the nonferrous metals and nonmagnetic steels. Inspection methods covered by this standard shall be of the following types. Type I - Dye penetrant (water washable) Type II - Dye penetrant (nonwater washable) Type III - Fluorescent penetrant (integral emulsification) Type IV - Fluorescent penetrant (post-emulsification) 2.0Surface Preparation: 2.1General - Surface of welds, castings, or wrought metals may be inspected without surface preparation or conditioning except as required to remove seals, slag, and adhering or imbedded sand. Blasting shall be accomplished by using angular or subangular cutting type sand, silicon carbide, or alumina grit. When blast peening, using steel shot, etc., is necessary before the penetrant inspection test, the blast peening shall be followed by blasting with angular or subangular cutting type sand, silicon carbide, or alumina grit, or by chemical cleaning. "As welded" surfaces, following the removal of slag, shall be considered suitable for liquid penetrant inspection without any grinding, provided the weld contour blends into the base metal without undercutting, and the contour and surface finish of the weld is in accordance with applicable specifications. 3.0Test Procedures: 3.1Protect Cleanliness - All materials being tested shall be cleaned by hot running water, by dipping in a solvent, or by swabbing with a clean lint-free cloth saturated with acetone. 3.2 The temperature of the penetrant and the part to be inspected shall be maintained between 50°F and 125°F. When inspection is necessary under conditions where the temperature of the penetrant and the inspection surface is outside the 50°F to 125°F range, the temperature shall be adjusted to bring them within this range. Due to the flammable nature of liquid penetrant inspection materials, the use of an open flame for heat purposes is prohibited.

NMP Unit 1 UFSAR UFSAR Revision 16 XVI-95 (EXHIBIT 8 Sh 2 of 4) November 1999 COMBUSTION ENGINEERING, INC. CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.3.1(a) DATE: April 27, 1960 SHEET: 2 of 4 3.3 The surface to be tested shall be thoroughly coated with penetrant by spraying, brushing, or immersion. The surface shall be kept wetted for the minimum time specified for the method employed: Penetrant Penetration Time All Applications Type I 30 Minutes Type II 10 Minutes Type III 30 Minutes Type IV 10 Minutes 3.4 The Type IV emulsifier shall be applied either by dipping or spraying the part. It should not be applied by means of a brush since stroking with a brush may remove the penetrant from shallow or scratch like discontinuities. After a suitable penetration time and emulsification period the surface film of the penetrant and emulsifier shall be removed from the part by employing a hot water spray not exceeding 120°F. Washing shall be checked under a black light to insure complete cleaning of all surfaces. Alternatively, the penetrant may be removed by use of the cleaner recommended by the manufacturer of the penetrant. 3.5 The penetrant of Types I and III shall be removed from all surfaces by swabbing with a clean lint-free cloth saturated with clear water or by spraying with water at moderate pressure. Alternatively the penetrant may be removed by wiping the excess penetrant from the test surface with a clean dry lint-free cloth followed by wiping the partially cleaned surface with an alcohol dampened clean cloth until all traces of the penetrant have been removed. 3.6 The Type II penetrant shall be removed from all test surfaces by wiping with clean lint-free dry rags using the cleaner recommended by the manufacturer. Excessive application of the cleaner shall be avoided to prevent the possibility of removing the penetrant from discontinuities, causing a decrease in the sensitivity of the test. It is best to slightly dampen a cloth with cleaner and wipe the part rather than flush the part with liquid cleaner. 3.7 The drying of test surfaces shall be accomplished by using circulating air, blotting with paper towels or clean lint-free cloth or normal evaporation. It is important that in the drying operation no contamination material such as oil from air nozzles or lint from rags be introduced into the surface which may cause misinterpretation during the inspection operation. 3.8 Dry developing powders shall be uniformly applied to surfaces resulting in a dust like appearance. Dry developers require a dry surface before application, or it will mat heavily on the wet surfaces. A short time shall be allowed for development of indications after the developing powder is applied. This time shall be about half as long as the time allowed for penetration. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-96 (EXHIBIT 8 Sh 3 of 4) November 1999 COMBUSTION ENGINEERING, INC. CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.3.1(a) DATE: April 27, 1960 SHEET: 3 of 4 3.9 Wet type developers shall be uniformly applied to surfaces by dipping, spraying or brushing. When using liquid type developers it is necessary that they continually agitate in order to prevent settling of solid particles dispersed in the liquid. Pools of wet developer in cavities on the inspection surface shall be avoided since these pools will dry to an excessively heavy coating in such areas resulting in the masking of indications. 4.0 Test Results: 4.1 General - All indications revealed by the penetrant are not necessarily defects since nonrelevant indications are frequently encountered in liquid penetrant inspection, generally due to failure to completely remove the excess penetrant from the surface being examined. At least 10 percent of such indications shall be removed to determine if defects are present. The absence of indications upon re-examination by liquid penetrant shall be considered to prove the indications nonrelevant in respect to actual defects. 4.2 Acceptance Standards for Welds 4.2.1 Examination of welds by liquid penetrant methods shall be made over an area including the weld and base metal extending for at least 1/2" on each side of the weld. 4.2.2 Surfaces examined by fluid penetrant methods shall be free of laps, fissures, cracks or other linear defects. 4.2.3 In any 6" length of weld and adjacent base metal examined, there shall be no indications greater than 1/32" diameter, as revealed by fluorescent penetrant examination, or 1/16" diameter as revealed by the dry penetrant. Nor shall there be more than six (6) indications whose sum of diameters is greater than three (3) times the maximum diameter specified herein. Any 6" length of weld shall be interpreted to denote the 6" length selected in the least favorable location with respect to the discontinuities disclosed by the inspection test. Indications of pinpoint porosity may be permitted if well dispersed, and if the pattern formed does not indicate that they are linearly disposed so as to promote formation of a crack or other continuous defect under stress. In linearly disposed porosity the average of the center-to-center distances between any one indication and the two adjacent indications shall not be less than 3/16". 4.3 Acceptance Standards For Weld Deposited Cladding 4.3.1 Examination of weld deposited cladding by liquid penetrant methods shall be made over a 4" diameter circular area. In lieu of 4" diameter circular area, a 3 1/2" square area may be used. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-97 (EXHIBIT 8 Sh 4 of 4) November 1999 COMBUSTION ENGINEERING, INC. CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.3.1(a) DATE: April 27, 1960 SHEET: 4 of 4 4.3.2 Surfaces examined by fluid penetrant methods shall be free of laps, fissures, cracks, or other linear defects. 4.3.3 In any 4" diameter circular area of weld deposited cladding examined, there shall be no indications greater than 1/32" diameter, as revealed by fluorescent penetration examination, or 1/16" diameter as revealed by the dye penetrant. Nor shall there be more than six (6) indications whose sum of diameters is greater than three (3) times the maximum diameter specified herein. Any 4" diameter circular area of cladding shall be interpreted to denote the 4" diameter circular area selected in the least favorable location with respect to the discontinuities disclosed by the inspection test. Indications of pinpoint porosity may be permitted if well disposed and if the pattern formed does not indicate that they are linearly disposed so as to promote formation of a crack or other continuous defect under stress. In linearly disposed porosity the average of the center-to-center distances between any one indication and the two adjacent indications shall not be less than 3/16". 4.4 Defect Removal 4.4.1 Only such defects need be removed and repaired as to render the weld acceptable to the requirements of this standard. Defective areas may be repaired by welding performed in accordance with an approved welding procedure. The areas containing defects shall be ground out to remove the defects. The ground out areas shall be reinspected to ascertain the complete removal of the defect. Completed repairs shall be reinspected by the method originally used. 5.0 Exception: 5.1 The acceptance criteria of Paragraph 4.3 shall not apply where the weld has been subjected to radiographic examination. Acceptability of radiographed welds shall be based on their radiographs compared to the applicable standard of acceptance. Defects defined in Paragraph 4.2 shall be removed or repaired regardless of radiographic test results. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-98 (EXHIBIT 9 Sh 1 of 2) November 1999 COMBUSTION ENGINEERING, INC. CORPORATE METALLURGICAL DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.4.3(b) DATE: June 9, 1960 SHEET: 1 of 2 SPECIFICATION FOR ULTRASONIC TESTING OF FORGINGS 1.0 Scope: 1.1 This specification provides for the method and technique for ultrasonic testing forgings. 1.2 Ultrasonic testing is an extremely sensitive tool capable of showing inclusions, small segregations, grain structure, differences, and small inhomogeneities which may be completely nonharmful for service. Cracks or laminations in metals are best revealed when the plane of the discontinuity is normal to the direction of the sound transmissions; for this reason complex shapes may place a limitation on use of ultrasonics where it is difficult to receive the transmitted sound for recording. Discontinuities revealed by ultrasonic inspection should be correlated, if feasible, with one or more of the nondestructive tests such as radiography or investigated by probing and/or sectioning in order to determine the acceptability of a forging. 2.0 Equipment: 2.1 Pulse reflection type equipment (Sperry UR Reflectoscope or equivalent) shall be used with the single crystal method for ultrasonic testing. 2.2 A suitable liquid couplant shall be used in sufficient quantities as to insure continuous contact. 3.0 Surface Condition: 3.1 The surface of the forging shall be equivalent to 250 RMS or better. It shall be clean, free of dirt and excessively rough or loose scale. 4.0 Definitions: 4.1 Initial pulse is the pulse observed to the left of the screen. 4.2 End reflection in longitudinal beam testing is the first back reflection from the surface of the forging opposite the crystal, that appears to the right of the initial pulse. 4.3 Indication denotes any reflection that may appear between the initial pulse and the end reflection. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-99 (EXHIBIT 9 Sh 2 of 2) November 1999 COMBUSTION ENGINEERING, INC. CORPORATE METALLURGICAL DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.4.3(b) DATE: June 9, 1960 SHEET: 2 of 2 5.0 Calibration and Test Procedure: 5.1 All forgings shall be tested at a frequency of 2.25, or one megacycle if necessary, using a crystal of one square inch area using the longitudinal beam technique. Ring and other hollow round forgings shall, in addition, be tested using an angle beam crystal having an area of one square inch. The ultrasonic test frequency shall be 2.25 megacycle or one megacycle except that final evaluation shall be made using one megacycle frequency. The depth of the calibration notch shall be 2 1/4% to 3% of the nominal section thickness or 3/4" (+0 - 3/16") whichever is smaller. The angle beam inspection need only be made in the circumferential direction in the case of ring forgings. 6.0 Test Results and Records: 6.1 Longitudinal beam technique - forgings containing one or more discontinuities which produce indications accompanied by a complete loss of the back reflection shall be subject to rejection. 6.2 Angle beam technique - forgings containing one or more discontinuities which produce indications exceeding in amplitude the indication from the calibration notch shall be subject to rejection. 6.3 If a forging has individual indications exceeding 12 1/2% or clusters of indications exceeding 5% of the end reflection, or of the calibrated notch amplitude, a chart of the forging shall be prepared and submitted to the purchaser. This chart shall show the approximate locations and magnitude of indications, accompanied by adequate notes. ` 6.4 If the forging producer and the purchasing agency are unable to agree on the interpretation of the ultrasonic indications, additional nondestructive tests or trepanning shall be performed to determine acceptability of the forging. 7.0 Purchaser Presentation: 7.1 Notification shall be given the purchaser at least 7 days prior to the time of the final ultrasonic testing. 7.2 A representative of the purchaser and of his customer may witness ultrasonic testing and may prepare the chart required by paragraph 6 if they so desire. All negotiations regarding the forging or the testing shall be conducted between C-E and the vendor. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-100 (EXHIBIT 10 Sh 1 of 3) November 1999 COMBUSTION ENGINEERING, INC. METALLURGICAL RESEARCH AND DEVELOPMENT DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.4.3(a) DATE: April 7, 1960 SHEET: 1 of 2 SPECIFICATION FOR ULTRASONIC TESTING OF PLATE MATERIAL 1.0 Scope: 1.1 This specification provides for the method and technique for ultrasonic testing of flat or shaped plate exceeding 3/8" thickness. 1.2 Ultrasonic testing to the requirements of this specification may be a provision of the purchase order, or it may be required by reference to this specification in a Material Purchase Specification. This specification also shall govern C-E shop inspection of plate when required by shop order. 2.0 Definitions: 2.1 Laminar Type Defect: A defect whose major dimension is in a plane approximately parallel to the major dimension or surface of the plate. 2.2 Inclusion Type Defect: Small discontinuities, inclusions or unwelded porosity which are scattered in relation to the thickness of the plate. 3.0 Equipment and Surface Conditions: 3.1 Pulse-reflection type equipment (Sperry UR Reflectoscope or equivalent) shall be used with the single crystal contact method of ultrasonic testing of plates over 3/8" in the thickness. Plates 3/8" and less in thickness will be tested by methods mutually agreed to by the manufacturer and purchaser. 3.2 Ultrasonic testing shall be performed using longitudinal wave 1 1/8" diameter, 2 1/4 Mc crystals. Deviations from this procedure may be requested of the purchaser. Crystals of other sizes and frequencies may be used for exploration or study of flaw indications. 3.3 The surface of plate to be tested shall be clean and free of dirt, excessive roughness or loose scale. 3.4 A suitable liquid sonic couplant shall be used in sufficient quantity and continuous sonic contact can be maintained. 4.0 Degrees of Testing: 4.1 Longitudinal wave testing for laminar discontinuities shall be performed on one surface of the plate being used. Scanning for defects shall be performed along parallel lines down on the plate, or indicated at the plate edges, at a spacing not greater than the crystal width. The extent of any indications appearing on the screen shall be investigated by searching locally over the area. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-101 (EXHIBIT 10 Sh 2 of 3) November 1999 COMBUSTION ENGINEERING, INC. METALLURGICAL RESEARCH AND DEVELOPMENT DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.4.3(a) DATE: April 7, 1960 SHEET: 2 of 2 5.0 Calibration: 5.1 Calibration sensitivity shall be established for longitudinal wave testing by adjustment of the instrument so that the first back reflection is approximately three-fourths of the screen height. 6.0 Test Results: 6.1 Plate with defects greater than 1 1/8" in diameter which eliminate the end reflection are subject to rejection. 6.2 A report shall be made to the purchaser, prior to shipment of plate, where laminar or inclusion type discontinuities are disclosed which reduce the back reflection by 50% or more; or where discontinuities are disclosed which produce traveling indications accompanied by a reduced back reflection. 7.0 Records: 7.1 A plan diagram of each plate tested which shows indications of the type referenced in paragraph 6.0 shall be prepared. This plan diagram will consist of marking such defects or areas in approximate actual location with the required dimensions and also the location of the mill heat number stampings. The plan diagram shall use the identifying classification and symbols given below insofar as is practicable. 7.1.1 Laminar defects which cause total loss of initial back reflection, and which provide a reflection, shall be indicated on the plan diagram, and identified by the letters "LT" (laminar defects causing total loss or back reflection). These defects shall be indicated as individual defects or, if large numbers are present in an area, as area defects. 7.1.2 Laminar defects which lower the normal back reflection by 50 percent or more, but less than 100 percent, and which provide a reflection shall be indicated on the diagram as "LP" (laminar defects causing partial loss or back reflection). 7.1.3 Inclusion type defects which cause loss of back reflection but which do not provide a reflection, shall be indicated on the diagram as "IT" (inclusion defects causing total loss of back reflection). 7.1.4 Discontinuities which cause traveling indications accompanied by a reduction in the back reflection shall be shown by explanatory notes. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-102 (EXHIBIT 10 Sh 3 of 3) November 1999 COMBUSTION ENGINEERING, INC. METALLURGICAL RESEARCH AND DEVELOPMENT DIVISION M&P SPEC. NO.: 2.4.4.3(a) CONT. SUPPL.: 164/264-1(a) DATE: April 9, 1964 1.0 Scope: This Contract Supplement shall modify M&P 2.4.4.3(a) for use on these specific contracts. 4.0 Degree of Testing: 4.1 Revise as follows: Longitudinal wave testing shall be performed on one surface of the plate. Scanning for defects shall be performed along parallel lines, spaced six inches apart, both transverse and parallel to the axis of the plate. A temporary layout of the required spacing may be made on the plate edges. The extent of any indications appearing on the screen shall be investigated by searching locally over the area. 5.0 Calibration: 5.1 Revise as follows: For longitudinal wave tests of plates four (4) inches or greater in thickness, the standard shall be a one-inch diameter, flat-bottom hole with a depth not greater than 10% of the material thickness. Plates less than four (4) inches thick shall use a hole diameter of 25% and a depth of 10% of the material thickness. 6.0 Test Results: 6.1 Revise as follows: For longitudinal wave tests, a defect which produces loss of back reflection in excess of the standard defect during movement of the transducer more than two inches in any direction, unless the loss of back reflection is due to two or more clearly separated indications, shall be unacceptable. Also, any echo indication that exceeds 50% of the reflection from the reference defect, regardless of back reflection, and that is continuous during movement at the transducer more than three inches in any direction shall be unacceptable. 6.2 Revise as follows: A chart shall be maintained of all defects accompanied by a 50% or greater loss of back reflection. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-103 (EXHIBIT 11 Sh 1 of 2) November 1999 COMBUSTION ENGINEERING, INC. NUCLEAR COMPONENTS DEPARTMENT M&P SPEC. NO.: 2.4.4.9(c) DATE: October 6, 1964 SHEET: 1 of 2 ULTRASONIC TESTING OF WELD DEPOSITED CLADDING PROCEDURE AND ACCEPTANCE STANDARD 1.0 Scope: 1.1 This specification provides the method, technique and acceptance for testing of weld deposited cladding for lack of bond. 2.0 Definitions: 2.1 Bond Area: The area of juncture or interface between the base material and the cladding material. 3.0 Equipment and Surface Condition: 3.1 Pulse-reflection type equipment (Sperry UR Reflectoscope or equivalent) shall be used with a single crystal. Either the wheel transducer, immersion or contact method may be utilized. 3.2 Ultrasonic testing shall be performed using a longitudinal wave 2 1/4 Mc crystal. Crystal areas shall not exceed one square inch for normal testing. Other sizes and frequencies may be used for exploration or study or flaw indications. 3.3 When an automatic or semi-automatic translated transducer is used, a visual alarm system shall be set to automatically indicate trace line patterns equal to or in excess to that produced for calibration. The alarm must trip automatically and be reset manually. 3.4 The surface of the cladding to be tested shall be clean, free of dirt and of approximately the same surface finish as that of the calibration standard. 3.5 A suitable liquid sonic couplant shall be used in the quantity necessary to maintain continuous sonic contact. 4.0 Calibration: 4.1 A calibration standard containing a flat bottom groove in clad plate 0.35 inch maximum width by at least one crystal width long shall be prepared in the following manner. Weld deposited corrosion resistant cladding shall be applied to one surface of a ferritic alloy block to a depth of 5/32" minimum. The flat bottom groove shall be obtained by machining away the base metal to within a short distance of the interface. This area shall then be chemically etched with nitric acid to completely remove the base metal and leave an "as deposited" clad surface. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-104 (EXHIBIT 11 Sh 2 of 2) November 1999 COMBUSTION ENGINEERING, INC. NUCLEAR COMPONENTS DEPARTMENT M&P SPEC. NO.: 2.4.4.9(c) DATE: October 6, 1964 SHEET: 2 of 2 4.2 The ultrasonic equipment shall be calibrated by adjusting to obtain satisfactory resolution from the flat bottom groove in the calibration standard. 5.0 Test Procedure: 5.1 The clad surface shall be inspected for lack of bond at intervals, 1.4 times the material thickness, but not greater than 12 inches, transverse to the direction of welding. Unbonded areas equal to or in excess of calibration will require additional scanning of the surrounding material until the boundary of the discontinuity is established. All testing shall be from the clad side of the material. 6.0 Acceptance Standard: 6.1 An unbonded condition producing a trace line pattern equal to or in excess of that produced for calibration, which is continuous during movement of the transducer more than three (3) inches in any direction is unacceptable. In addition, unbonded areas producing patterns equal to or in excess of calibration, but which are less than three inches long, shall be unacceptable if separated by less than one (1) inch. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-105 (EXHIBIT 12 Sh 1 of 3) November 1999 COMBUSTION ENGINEERING, INC. METALLURGICAL RESEARCH AND DEVELOPMENT DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.2.1(b) DATE: June 29, 1960 SHEET: 1 of 3 PROCESS SPECIFICATION FOR MAGNETIC PARTICLE TESTING DRY POWDER PROD METHOD 1.0 Scope: 1.1 This process specification provides for the method and standard of acceptance for dry powder prod method of testing of ferritic materials. 1.2 Magnetic particle testing is used for determining the presence of discontinuities at or near the surface of ferromagnetic metals. This test is particularly useful for locating cracks on surfaces of steel and for examining chipped or ground cavities prior to welding. Laminations on edges of wrought plate may also be detected by the magnetic particle test method. 1.3 Magnetic particle testing shall be performed at the various stages of production that is required by the applicable C-E General Welding Specification and customer contract specifications. Records shall be maintained of each inspection when performed to satisfy specification requirements. 1.4 Magnetic particle inspection shall include base metal for at least 1/2" on each side of the weld where possible. 2.0 Surface Preparation: 2.1 General - Surface of castings or wrought metals may be inspected without surface preparation or conditioning except as required to remove scale, slag, and adhering or imbedded sand. 2.1.1 Preparation for Test - Surfaces of parts shall be dry and free of oil or any other material which might interfere with the formation of interpretation of magnetic particle patterns or indications. Small openings and oil holes leading to obscure passages shall be plugged with a suitable nonabrasive material which is readily removable. Unless otherwise required by contract specification, "as-welded" surfaces following the removal of slag, shall be considered suitable for magnetic particle testing without grinding, provided the weld beads blend into each other and into the base metal without undercut, overlap, or abrupt ridges or valleys. The edge of the crown of welds shall be faired out to blend with the adjacent base metal. Caution should be exercised to avoid undergrindings. 2.2 The surface of the piece being tested shall be free of scale or rust in the area of prod contact points to provide good electrical contact and avoid arcing or overheating of the contact area. Prods should be kept free of iron pick-up by frequent filing. Local areas of the metal being tested which have been subjected to arcing shall be ground to clean metal. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-106 (EXHIBIT 12 Sh 2 of 3) November 1999 COMBUSTION ENGINEERING, INC. METALLURGICAL RESEARCH AND DEVELOPMENT DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.2.1(b) DATE: June 29, 1960 SHEET: 3 of 3 6.0 Standards of Acceptance: 6.1 Surfaces examined by magnetic particle testing shall be free of laps, fissures, cracks or other linear defects of crack-like nature. In-line porosity which appears as a linear accumulation of magnetic particle powder shall be removed. 7.0 Repairs: 7.1 Only such defects need be removed and repaired as to render the weld acceptable to the requirements of this standard. Defective areas may be repaired by welding performed in accordance with the procedure approved for such repairs on the individual contract. Completed repairs shall be reinspected by the method originally used. Minor cavities resulting from the removal of shallow discontinuities that do not reduce the thickness to below minimum design requirements, and do not adversely affect the strength, machinability or other service requirements of the part, need not be weld repaired but shall be blended into the surrounding area. 7.2 Defects which do not require repair welding shall be blended out to a bottom radius at least three times the depth of defect cavity. The bottom radius shall extend to the surface from which the defect originates. Inter-section at the surface from which the defect originates shall be accomplished with a smooth radius. The above blend-out shall extend completely around the defect. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-107 (EXHIBIT 12 Sh 3 of 3) November 1999 COMBUSTION ENGINEERING, INC. METALLURGICAL RESEARCH AND DEVELOPMENT DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.2.1(b) DATE: June 29, 1960 SHEET: 2 of 3 3.0 Powder Application: 3.1 The test powder shall normally be Magnaflux Corporation No. 1 (gray). No. 3A (black) or No. 8A (red) may also be used. 3.2 Preferred application of testing powder shall be with a mechanical blower. 4.0 Magnetization: 4.1 Testing shall be performed with direct current, or half wave or full wave rectified current. Alternating current shall be used for testing the first layer of fillet welds, and also may be used to assist in the evaluation of indications. An initial surge of approximately twice the testing current may be employed when this is provided in the power source equipment. 4.2 Normal testing of large surfaces shall provide for inspection of approximately 5" square areas with prods so located as to give coverage of the perimeter and diagonals of the area. 4.3 Normal testing current shall be approximately 600 amps for 6" prod spacing. This should be reduced to approximately 400 amps for 4" spacing and increased to 800 amps for 8" spacing. Prod specimens closer than 4" may give false readings due to the magnetic field around the prod. Spacing greater than 8" may be used for inspecting plate surfaces for surface defects provided the magnetizing current is increased proportionally. These values shall be used for section thicknesses exceeding 3/4". For thinner sections the current values should be reduced by 15% to 25%. 5.0 Interpretation of Indications: 5.1 The powder shall be applied while current is passing through the piece being tested in such a manner that a light, uniform, dust-like coating settles on the surface. The application of powder shall be stopped and the surface under test examined with the magnetizing current on while blowing with a mild air steam with the nozzle approximately 2" from the surface. (This would be the air flow obtained from 50 feet of 5/16" hose with a gage reading of 1.75 psi. The magnaflux powder dispenser and blower is calibrated to provide a correct air flow). The powder pattern should again be observed while passing the air current over the surface after the current is turned off. 5.2 Operator skill must be developed in the interpretation of powder indication patterns especially in the evaluation of subsurface defects vs. unfused roots on partial penetration welds. This can best be developed through observation by him of excavations made to explore indications. It is important to provide to the magnetic particle testing operator the exact details of nonthrough penetrating welds to assist him in the evaluation of powder patterns. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-108 (EXHIBIT 13 Sh 1 of 2) November 1999 COMBUSTION ENGINEERING, INC. METALLURGICAL RESEARCH AND DEVELOPMENT DIVISION CHATTANOOGA, TENNESSEE M&P SPEC. NO.: 2.4.6.1(a) DATE: May 12, 1960 SHEET: 1 of 1 HYDROSTATIC TESTING OF NUCLEAR PRESSURE VESSELS 1.0 Scope: This specification provides special methods required for the hydrostatic leak testing of nuclear pressure vessels to assure structural integrity and freedom from leaks. 2.0 General: 2.1 Hydrostatic tests shall be performed in accordance with the applicable section of the ASME Code for Unfired Pressure Vessels, Section VIII, except where modified by this specification, or fabrication drawings. 3.0 Conditions for Test: 3.1 The test pressure and minimum vessel temperature shall be specified on fabrication drawings. 3.2 Reactors and the primary side of steam generators and miscellaneous stainless or stainless clad vessels may be tested with city tap water where final cleaning operations on internal surfaces have not been completed. 3.3 Steam generators shall be tested with distilled or demineralized water on both the primary and the secondary sides. 3.4 The vessel or chamber shall be held at a test pressure for the period of time specified by Section VIII of the ASME Code for Unfired Pressure Vessels. Where customer specifications are different from code requirements the holding time shall be given on fabrication drawings. The pressurizing pump shall be disconnected during the holding time at pressure. 3.4.1 The vessel surface accessible shall be visually examined for leaks while the vessel is pressurized. 3.4.2 The cause for a drop in the gage pressure during the holding time, exceeding that which would be caused by a lowering of the temperature, shall be determined and corrected. After correction the pressure test shall be again applied. 3.4.3 No hammer blows shall be employed during the test. 4.0 Drying of Steam Generators: Following the hydrostatic testing of the secondary side of steam generators, the test water shall be drained and drying shall be performed by evacuation to 100 microns in accordance with the requirements of M&P Process Specification M&P No. 5.5.3.1 for Cleaning of Nuclear Components. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-109 (EXHIBIT 13 Sh 2 of 2) November 1999 ENGINEERING DATA SHEET NUMBER 09596-3 SUBJECT C. E., Inc. M&P Spec. No. 2.4.6.1(a) Hydrostatic Testing of Reactor Pressure Vessels PROJECT Niagara Mohawk DATE February 26, 1964 EQUIPMENT Reactor Vessel VFF No. 1236-14 CONTROL NO. 205-09596 INITIATING ENGINEER T. F. Robinson (sig) SHEET 1 OF 1 Refer to C. E., Inc. M&P Spec. No. 2.4.6.1(a) - Hydrostatic Testing of Nuclear Pressure Vessels, sheet 1 of 1 paragraph 3.1 Comment: GE Co. APED interprets this to mean that hydrostatic test pressure and minimum metal temperature required for the vessel during hydro testing will be specified along with other Code Information on a prominent drawing of the set of drawings that are part of this order. With the above comment the subject specification is approved. NMP Unit 1 UFSAR UFSAR Revision 16 XVI-110 November 1999 F. CONTAINMENT DESIGN REVIEW This section contains a summary of design analysis calculations utilized in the construction of the primary containment. Nine Mile Point Unit 1 UFSAR CHICAGO BRIDGE & IRON COMPANY 901 WEST 22NO STREET, OAK BROOK, ILLINOIS 90!521 July 7, 1967 CERTIFIED STRESS REPORT NINE MILE POINT CONTAINMENT VESSEL As a Professional Engineer registered in the State of .Illinois with License No. 62-21203, I do hereby certify that the stress .report for the subject vessel has been reviewed by me and satisfies the requirements of the Technical Specification and the ASME Pressure Vessel Code for Nuclear Vessels (Section III, Class B) and I do further certify that I am experienced in pressure vessel design as required under Section III Par. N-142 of the Vessel Code. Area Code: 312 654-1700 UFSAR Revision 16 XVI-111 (EXHIBIT 14 Sh 1 of 10) November 1999 Nine Mile Point Unit 1 UFSAR A TELIDTNE COMPANY SUMMARY OF DESIGN .ANALYSIS CALCULATIONS-1.0 Code Approval Calculations 1.1 Basic Structure THEDTNE MATHIALS HSEAICH 303 HAI HILL IOAD WALTHAM. MASSACHUSETTS 0215* 16171 eff.JUO The basic structure discussion is intended to cover all tions of the pressure suppression containment other than those considered in Section 1.2. This permits a common cussion of the bellows which occur in the nozzle extensions as well as those which occur in the suppression chamber. Section 1.2, then, considers all but the bellows portion of the nozzle extensions. l.l.2 Method Stress analyses have been performed by the manufacturer for each of the components considered herein. In the case of the basic containment structure, including the suppression chamber bellows, an ASHE Stress Report was prepared and certified by the Chicago Bridge and Iron Company. In the case of the nozzle extension bellows, an analysis was provided by Temp Flex to the Niagara Mohawk Power Corporation. These analyses have been reviewed by TMR personnel for the purpose of establishing whether or not the analyses are adequate to demonstrate that the design complies with the requirements of Section III of the ASME Boiler and Pressure Vessel Code. l.l.3 Conclusions Based upon the review conducted, TMR has concluded that the existing analyses are not adequate to establish that the sign complies with the requirements of Section III. of the ASHE. Boiler and Pressure Vessel Code. However, TMR has stated an opinion that the design is adequate for the intended ment function. This apparent conflict results from differing interpretations of the ASME Section III requirements by the parties involved, and is discussed in more detail in Section l.1.4. INGIHHU AND MITAUUIOllTI J -T LDStiLS AND AllOCIATn, INC. & NIW llHGIAHD MATHIALS l.AIOIATOIT, INC. UFSAR Revision 16 XVI-112 (EXHIBIT 14 Sh 2 of 10) November 1999 0 0 0 Nine Mile Point Unit 1 UFSAR 1.1.4 Discussion a. Dryweli & Suppression Chamber TMR is in agreement with the Stress Report methods used by the Chicago Bridge and Iron Company in all respects except for the treatment given to secondary stresses in regions removed from the sand transistion zone and moved from regions of local loading. Of major consequence with respect to this disagreement is the type of used for the main closure flange, the flanges associated with the equipment door and other similar openings, and the suppression chamber bellows. This last item is considered in Section 1.1.4 b, of this report. The regions where there are disagreements with respect to analytical methods are those regions of structural continuity where the stresses are primarily due to pressure. CB&I interpretes the Code to pen:iit the use.of the formulas and rules of ASMI:: Section VIII so long as there are no substantial loads other than pressure. With this pretation, the stress limits of paragraph N-414, which require a detailed analysis, have been considered to apply only when other substantial mechanical or thermal loads are present. TMR interpretes the Code as requiring a tailed analysis and conformance with N-414 for all loadings. TMR concurs with the CB&I opinion that their tion provides design procedures consistent with the service requirements of containment vessels and will support the proposal made by CB&I (in a letter of August 2, 1967 to the Code Secretary) to make appropriate changes in the Code wording. In addition, CB&I and TMR are in agreement that the most significant discontinuity stresses occur in the sand transistion zone and that a proper analysis has been made of this effect. This paragraph indicates why TMR has stated their opinion that the design is adequate for the intended service. b. Bellows The utilization of bellows in containment vessels is thorized by Code Case 1330 (previously Case 1276N)

• Both the inquiry and the reply refer to expansion joints "under the rules of Section III and Case 1177". The problem is similar to that discussed in the previous section, in that TELEDYNE MATERIALS RESEARCH COMPANY "\J -engineers and metallurgists --boston -UFSAR Revision 16 XVI-113 (EXHIBIT 14 Sh 3 of 10) November 1999 Nine Mile Point Unit 1 UFSAR Case 1177 was written to permit bellows in Section VIII vessels.and contains design rules consistent with Section VIII. How, then, are the words referrring to Section III to be interpreted? CJS&I applies an interpretation that the design rules of Case 1177 are applicable, THR applies an interpretation that requires performance of a detailed analysis and conformance with N-414. A diatinction must now be made between the suppression chamber bellows (the responsibility of CB&I) and the nozzle extension bellows (supplied by Temp Flex). The suppression chamber bellows are an integral part of the suppression chamber and experience loadings similar to those ienced by the suppression chamber. Therefore, TMR is willing to state that the design of the suppression chamber bellows is adequate for the same regions as were mentioned in the last paragraph of the preceding section. In contrast, the nozzle extension bellows are subjected to stress during normal operation as they relative motion between the drywell and the piping. This is a fatigue problem and ASME Section III has recently been clarified to make it obvious that such members require a complete analysis and that the rules of N-414 must be satisfied. Such an analysis has not been provided by Temp Flex. TMR is willing to agree with the adequacy of the design of these bellows for the intended service only because a double barrier is provided and periodic testing of these bellows is planned. TELEDYNE MATERIALS RESEARCH COMPANY ,.\J -engineers and metallurgists --boston -UFSAR Revision 16 XVI-114 (EXHIBIT 14 Sh 4 of 10) November 1999 0 0 0 Nine Mile Point Unit 1 UFSAR l.2 Penetrations l.2.l Analysis Performed Two separate analyses were performed: Stresses in the guard pipe resulting from imposed piping system deflections*, and stresses in the penetration sleeve sulting from piping system thermal transients and internal pressure. The specific piping system thermal transients considered for each nozzle extension are as follows: a. Main Steam -lOOF to 550F in approximately 4.5 hours. b, Main Feedwater -lOOF to 300F instantaneously, then 300F to 368F in one hour. c. l::mergency Cooling -l20F to 550F instantaneously. This condition results from a stagnant leg. of 120F water in the system piping at the containment tration, resulting in the nozzle extension being thermal at l20F. A shot of 550F hot steam is released from the condenser when the emergency condensate valve at the reactor is opened. d. Core Spray -540F to 140F instantaneously. This dition results from the assumption that reactor valve leakage occurs and the containment penetration sleeve is isothermal at 540F. The core spray valve is then opened, admitting l40F water to the system. e. Reactor Cleanup -lOOF to 532F instantaneously. Ihis condition results from the cleanup system being at lOOF prior to startup and the penetration sleeve fore being isothermal at this temperature. The reactor cleanup valve is then opened, admitting 532F water to the system. f. Shutdown Cooling -lOOF to 350F instantaneously. l'his results from the system being isothermal at lOOF prior to shutdown. The shutdown cooling valve is then opened admitting 350F water to the system. MA TERI A LS RESEARCH COMPANY -engineers dnd mctdllurgish --boston -UFSAR Revision 16 XVI-115 (EXHIBIT 14 Sh 5 of 10) November 1999 Nine Mile Point Unit 1 UFSAR Stresses in the penetration guard pipe sulting from 62 psi internal pressure are also found. 1.2.2 Procedure The general design criteria used in this report is as prescribed in the Code under N-1312 of Article 13. The basic techniques used in performing the analysis are: a. due to icposed pipins deflections are culated using simple beam theory using the equations of Reference l. b. Metal temperature distribution in the nozzle sions are found using LION-L, "Temperature tions for Arbitrary Shapes and Complicated Boundary Conditions." (2) is a digital computer program which will solve three-dimensional transient and state temperature distribution problems. The input consists of geometry, physical properties, and boundary conditi'ons as a function of title. In addition to solving problems of heat conduction in a structure, LIOH can handle forced convection, free convection, and radiation or a cOlllbination of these at the surface of the structure. The output consists of complete nodal temperature distributions along with surface fluxes and surface heat transfer coefficients. For the specific cases analyzed, thickness gradients were not required since the stresses in the penetration sleeves are a fw1ction of relative bulk temperature differences. The insulated surfaces were assumed to be ideal and no heat transfer was taken across these boundaries. Temperatures were found at a number of times throughout the transient until a steady-state condition was reached. Steady state was specified as a change in temperature of any node less than c. Stresses in the nozzle extension structure due to tem-perature differences were found using SOR-II, "A Pro-(3) gram to Perform Stress Analysis of Shells of Revolution." SOR-II analyzes general surfaces of revolution. The program solves for the forces, deflections, stresses, and strains in shells of revolution with variable thicknesses and elastic moduli. The axisymmetric loadings considered include arbitrary distributions of normal, tangential, and moment surface loadings, as .... MATERIALS RESEARCH COMPANY -engineers and metallurgists --boston -UFSAR Revision 16 XVI-116 (EXHIBIT 14 Sh 6 of 10) November 1999 0 0 0 Nine Mile Point Unit 1 UFSAR well* as edge forces and thickness directions. The additional effects of misalignment, line loads, and elastic supports at the shell sections are considered. For the specific cases analyzed, the guard pipe and valve nozzle were applied as elastic supports to the remaining penetration structure (system pipe, valve flange, and spool piece). The deflections and rotations of the valve flange at the point of guard pipe attachment were then applied to the guard pipe. This procedure sulted in stresses throughout the structure. 1.2.3 Results of Analysis a. Stresses due to imposed piping system deflections. (1) The allowable stress is 1.1 times the value for Al06-GR.B given in Section VIII of the ASML, Boiler and Pressure Vessel Code. sallowable al.l (15,000) = 16,500 psi (2) Stresses (4,5) Penetration Total Stress (psi) Main Steam 2,000 Main Steam 2,000 Emergency Cooling 4,600 Emergency Cooling 3,600 Feedwater 3, 300 Feedwater 3, 300 Shutdown 13,600 Shutdown 10,500 Cleanup 9, 000 Cleanup 3,000 Core Spray 5,000 Core Spray 15,200 Emergency Cooling Return 9,800 Emergency Cooling Return 8,100 .... MATERIALS RESEARCH COMPANY -r-ngineers and metallurgists --boston -UFSAR Revision 16 XVI-117 (EXHIBIT 14 Sh 7 of 10) November 1999 Penetration Main Steam Feedwater Emergency Condensate Core Spray Reactor Cleanup Shutdown Cooling UFSAR Revision 16 Nine Mile Point Unit 1 UFSAR b. Stresses due to piping system thermal transients. The allowable stress in the guard pipe and spool piece is 3Sm*49,500 psi; and in the flange 3SQ for 304 stainless steel
• 3(1.1 x 17,500)
• 57,750 psi. It should be noted that in accordance with (rn) (:) of the Code, the stress due to a radial gradient is classified as a stress anu is con:;idered in fatigue only. This is due to the almost complete suppression of the differential expansion and subsequent lack of significant tortion. This is the situation occurring in the valve flange for all penetrations; that is, the in plant! stresses are considered only in fatigue. Primary + Secondary Allowable Intensity (psi) (psi) 22,400 49,500 19,500 49,500 54,000 57,750 45,500 49,500 47,500 49,500 29,000 49,500 c. Stresses due to internal pressure. The design pressure of the containment vessel is 62 psi. This pressure was applied to the feedwater netration guard pipe to the effccl of internal pressure on the guard pipe stresses. JS As assW11ed, the stresses due to pressure glibible and analyses of other nozzles were not performed. The results of the feedwater are summarized here and are considered generally applicable to all nozzles. (l) General Primary Membrane Stress Intensity m The allowable is Sm = lb,500 psi. The general primary membrane stress is 1,500 psi. MATERIALS RESEARCH COMPANY -engineers dnd metallurgists --boston -XVI-118 (EXHIBIT 14 Sh 8 of 10) November 1999 0 0 0 Nine Mile Point Unit 1 UFSAR (2) Local Membrane Stress Intensity The allowable is l.5 Sm= 24,750 psi. The local membrane stress intensity is 4,458 psi. d. Fatigue evaluation. A fatigue evaluation is in accordance with the rules of N-415.2 of the Code. The maximum allowable cumulative usage factor (U) is 1.0. A summary of the results of this evaluation is provided in the following table. Penetration Steam Feedwater Reactor Cleanup Shutdown Cooling l:.mergency Condensate Cu:mnulative Usage Factor 0.333 O.t>77 0.114 0.238 0.007 MATERIALS RESEARCH COMPANY ,... -engineers and mE:'tallurgists --boston -UFSAR Revision 16 XVI-119 (EXHIBIT 14 Sh 9 of 10) November 1999 Nine Mile Point Unit 1 UFSAR References Section l.2.2 (Procedure) l. Raymond J. Roark, Forculas for Stress and Strain, First tdition, McGraw-Hill Book Company, Inc., New 2. J. R. Schmid, G. L. Lechliter, w. W. Fischer, "LION, Temperature Distributions for Arbitrary Shapes and Complicated Boundary Conditions". General Electric Company Report KAPL-M-6532. 3. J. A. Mirabal, D. G. Dight, "SOR-II -A Program to Perform Stress Analysis of Shells of Revolution" 1 General Electric Company Report MPL-M-EC-19. Section 1.2.3 (Results of Analysis) 4. Design of Piping Systems, The M. W. Kellogg Company, Second tdition, John Wiley b Sons, Inc., New York. 5. Norman Jones, "On the Design of Pipe Bends", Nuclear Engineering and Design, 4 (1966) 399-405, North-Holland Publishing Company, Amsterdam. MATERIALS RESEARCH COMPANY -c-ngincers and metallurgists --boston -L------UFSAR Revision 16 XVI-120 (EXHIBIT 14 Sh 10 of 10) November 1999 0 0 0 NMP Unit 1 UFSAR Section XVI XVI-121 Rev. 25, October 2017 G. USAGE OF CODES/STANDARDS FOR STRUCTURAL STEEL AND CONCRETE The analysis and design modifications/changes and/or additions to the plant structures may be done using AISC Manual of Steel Construction, 8th Edition, and ACI 349-85 for structural steel and concrete components, respectively, in lieu of AISC Manual of Steel Construction, 6th Edition, and ACI 318-63 for Structural Steel and Concrete Components. However, the requirements from different editions of the Code (e.g., design requirements from one edition with material requirements from another edition) shall not be mixed unless the differences in the requirements of the Code editions are reconciled. Any other codes/standards can also be used provided these codes/standards are reconciled to the above-mentioned AISC and ACI Codes.

NMP Unit 1 UFSAR Section XVI XVI-122 Rev. 25, October 2017 H. REFERENCES 1. Dames & Moore, "Design Acceleration Spectra Curves for Niagara Mohawk Project," San Francisco, July 14, 1964. 2. General Electric, "Nine Mile Point Reactor Building, Dynamic Analysis North-South Direction," March 5, 1965. 3. John A. Blume & Associates, Engineers, "Seismic Analysis of Reactor Pressure Vessel for the Niagara Mohawk Reactor Project," May 17, 1965. 4. General Electric Nuclear Energy Report No. GE-NE-B13-01739-03, Revision 0, Class III, "Nine Mile Point 1 Nuclear Power Station Seismic Analysis Core Shroud Repair Modification," December 1994. 5. MPM-USE-129215, Elastic-Plastic Fracture Mechanics Assessment of Nine Mile Point Unit 1 Beltline Plates for Service Level A and B Loadings. 6. MPM-USE-293216, Elastic-Plastic Fracture Mechanics Assessment of Nine Mile Point Unit 1 Beltline Plates for Service Level C and D Loadings. 7. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Reports MPM-USE-129215 and MPM-USE-293216 on Upper Shelf Energy Equivalent Margins Analysis. 8. BWROG Vessel Internals Project (VIP) Core Shroud Design Criteria, Revision 1, September 12, 1994. 9. R. J. Roark, Formulas for Stress and Strain, 270, Case 6, New York, 1954. 10. P. P. Bijlaard, Welding Research Bulletin No. 34, Welding Research Council, New York, 1957. 11. F. J. Moody, "Maximum Flow Rate of a Single Component, Two-Phase Mixture," GE APED-4378, Nine Mile Point Nuclear Station Preliminary Hazards Summary Report, First Supplement. 12. R. G. Keshock and C. E. DeBogdan, "Leak Rate Testing of NASA Plum Brook Reactor Containment Vessel," NASA UTN-D-1731, July 1963. NMP Unit 1 UFSAR Section XVI XVI-123 Rev. 25, October 2017 13. United States Atomic Energy Commission, Division of Reactor Development, "Nuclear Reactors and Earthquakes," Washington, D.C. 14. John M. Biggs, "Introduction to Structural Dynamics" (New York). NMP Unit 1 UFSAR Section XVI XVI-124 Rev. 25, October 2017 15. Charles H. Norris, et al, "Structural Design For Dynamic Loads" (New York). 16. M. W. Kellogg Company, "Design of Piping Systems" (New York: John Wiley and Sons, Inc.). 17. L. H. Chen, Transactions of ASME (December 1959), 608-612 "Piping Flexibility Analysis by Stiffness Matrix." 18. Theory," NRL Report 6002 (November 1963). 19. G. Braun, "Heat Transmission by Condensation of Steam on a Spray of Water Drops," International Conference, Heat Transfer, London, 1951. 20. F. Syenitzer, "The Evaporation or Cooling of a Liquid Drop Braked by Air," International Conference, Heat Transfer, London, 1951. 21. "Report on Use of H. H. Robertson Steel Roof and Floor Decks as Horizontal Diaphragms," S. B. Barnes and Associates, 1963. 22. Whitman and Richart, "Design Procedures for Dynamically Loaded Foundations," Proc. ASCE, SM6, November 1967. 23. H. C. Martin, "Introduction to Matrix Methods of Structural Analysis," McGraw-Hill, 1965. 24. Hurty and Rubinstein, "Dynamics of Structures," Prentice-Hall, Inc., 1965. 25. R. W. Clough, "Dynamic Effects of Earthquakes," Proc. ASCE, Vol. 86, ST4, April 1960. 26. "The Agadir, Morocco Earthquake," Chapter 4, American Iron and Steel Institute, 1962. 27. Blume, Newmark, and Corning, "Design of Multistory Reinforced Concrete Buildings for Earthquake Motions," Portland Cement Association, 1961. 28. Letter from R. R. Schneider (NMPC) to B. H. Grier (NRC), Final Response to IE Bulletin 79-02, December 10, 1979. 29. TID-7024, United States Atomic Energy Commission, 1963. NMP Unit 1 UFSAR Section XVI XVI-125 Rev. 25, October 2017 30.Letter from P. D. Raymond to A. Giambusso, Pipe Whip Analysis, June 29, 1973. 31. Fawbush, Miller & Starrett, "An Empirical Method of Forecasting Tornado Development," Bulletin, AMS, Fig. 4, 1951. NMP Unit 1 UFSAR Section XVI XVI-126 Rev. 25, October 2017 32. Spohn, et al, "Tornado Climatology," Monthly Weather Review, Washington, DC, Vol. 90, pp 398-406, 1962. 33. Flora, Tornadoes in the United States, University of Oklahoma Press, 1954. 34. Thom, "Tornado Probabilities," Monthly Weather Review, Washington, DC, Vol. 91, pp 730-36, 1963. 35. NRC Safety Evaluation, "Alternative Repair of the Core Shroud Vertical Weld," April 30, 1999. 36. NRC Safety Evaluation, "Supplemental SE Regarding Alternative Repair of the Core Shroud Vertical Welds," May 24, 1999. 37. BWRVIP-78, "BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan," Final Report, December 1999. 38. BWRVIP-86-A, "BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan," Final Report, October 2002. 39. Letter from USNRC to C. Terry (BWRVIP), "Safety Evaluation Regarding EPRI Proprietary Reports 'BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan (BWRVIP-78)' and 'BWRVIP-86: BWR Vessel and Internals Project, BWR Integrated Surveillance Program Implementation Plan'," dated February 1, 2002. 40. NRC Letter to NMPNS dated November 8, 2004, "Nine Mile Point Nuclear Station Unit Nos. 1 and 2 - Issuance of Amendments Re: Implementation of the Reactor Pressure Vessel Integrated Surveillance Program (TAC Nos. MC1758 and MC1759)." 41. NER-1M-093 Rev. 0, "Evaluation of Reactor Internals Pressure Differences at the Faulted Interlock Point for Low Power/High Flow Conditions." NMP Unit 1 UFSAR Section XVI XVI-127 Rev. 25, October 2017 TABLE XVI-1 CODE CALCULATION SUMMARY Part Section of ASME Code Reqd. Thk. or Reinforcing Area Actual Thk. or Reinforcing Area a. Bottom Head with Control Rod Drive Penetration I (par. 195) 8.750 in 33.8 in2 8.750 in min. 41.0 in2 b. Bottom Head I (par. 195) 3.435 in 8.750 in min. c. Vessel Shell I (par. 180) 7.092 in 7.125 in min. d. Vessel Flange VIII 16.03 in 16.28 in e. Closure Flange VIII 28.25 in 28.50 in f. Top Head I (par. 195) 3.4975 in 4.312 in g. Nozzles Recirculation Inlet Outlet Steam Outlet Feedwater Emergency Cooling Core Spray CRD Hydraulic Return Vent Top Head 6-In Instrument Top Head 2-In Instrument 1-In Instrument Core P Drain I (par. 268) 89.954 in2 244.89 in2 198.81 in2 75.463 in2 70.144 in2 44.437 in2 29.23 in2 15.24 in2 24.67 in2 21.28 in2 14.06 in2 12.02 in2 5.97 in2 142.66 in2 246.43 in2 220.64 in2 78.474 in2 70.327 in2 46.165 in2 29.967 in2 15.59 in2 24.73 in2 21.85 in2 14.781 in2 91.26 in2 17.38 in2 NMP Unit 1 UFSAR Section XVI XVI-128 Rev. 25, October 2017 TABLE XVI-2 STEADY-STATE - (100% FULL POWER NORMAL OPERATION) PERTINENT STRESSES OR STRESS INTENSITIES Part Material Allowable(1) Stress or Stress Intensity (psi) Governing Stress(2) or Stress Intensity (psi) Vessel Support SA336 Mod Code Case 1236 20,000 60,000 -7,600 20,200 -35,620 Bottom Head SA302B 20,000 60,000 7,800 9,900 Cyl. & PH Junction SA302B 20,000 60,000 8,200 8,800 Shroud Support Cone SB168 20,000(7) 60,000 15,340(8) 18,650(8) Closure Flange to Shell SA336 Mod Code Case 1236 20,000 60,000 10,700 43,500 Vessel Flange to Shell SA336 Mod Code Case 1236 20,000 60,000 15,900 40,600 Control Rod Drive Stub Tube SA312 TP304 15,033 45,100 44,100(3) Feedwater Nozzle SA336 Mod Code Case 1236 20,000 60,000 57,120(4) Core Spray Nozzle SA336 Mod Code Case 1236 20,000 60,000 46,840(4) Recirculation Inlet Nozzle SA336 Mod Code Case 1236 20,000 60,000 56,660(5) Recirculation Outlet Nozzle SA336 Mod Code Case 1236 20,000 60,000 50,400(5) Recirculation Nozzles Stainless to Carbon Steel Junction SA182 316NG 15,000 45,000 57,610(5,9) Studs SA540 Grade B23 or B24, Class 3 (4340) 36,325 72,650(6) 50,900 NMP Unit 1 UFSAR Section XVI XVI-129 Rev. 25, October 2017 TABLE XVI-2 (Cont'd.) (1) First listed allowable stress is the Code allowable stress Sm. The second listed stress is 3 Sm, the GE vessel specification primary plus secondary intensity limit. (2) First stress is the average membrane stress through the section due to pressure and mechanical loads. The second listed stress is the calculated primary plus secondary stress intensity for the 100-percent full power normal operation unless otherwise stated. The third stress is the primary and secondary stress intensity during automatic blowdown. (3) Calculational methods yield only primary plus secondary stresses. This case includes earthquake plus a stuck control rod. (4) Calculational methods yield only primary plus secondary stresses. (5) Result obtained from thermal transient analyses performed under Design Change N1-00-010. (6) Allowable stress based on 1/3 of the minimum yield point at temperature for the particular stud material. The second listed allowable is only 2 Sm, or 2/3 the minimum yield point at temperature, and is used as the primary plus secondary stress limit for studs. (7) The allowable stress value of 20.0 ksi for the Inconel 600 shroud support cone was based on ASME Code, Section VIII, which is similar to current-day ASME Section III, Class 2/3 allowable stresses. Since the shroud support cone was fabricated as part of the reactor vessel, the Class 1 allowable of 23.3 ksi for Inconel 600 was used in the "Nine Mile Point Unit 1 Shroud Repair Hardware Stress Analysis," dated January 1995 (GE Nuclear Energy Report No. GE-NE-B13-01739-04). (8)` The governing stresses in the shroud support cone have been revised to reflect the attachment of tie-rod assemblies to the cone. The tie-rod assemblies which are bolted to the cone at four separate azimuths were installed under Shroud Repair Modification N1-94-003. For complete details of the stress analyses performed for the shroud repair modification, refer to GE Nuclear Energy Reports GE-NE-B13-01739-04 and 24A6426. (9) Complies with ASME Code requirement for exceeding 3 Sm allowable because fatigue limits with Code-required penalty factors are met. NMP Unit 1 UFSAR Section XVI XVI-130 Rev. 25, October 2017 TABLE XVI-3 LIST OF REACTIONS FOR REACTOR VESSEL NOZZLES This table represents historical information that demonstrated conformance with applicable codes at the time of the original FSAR submittal. Subsequent design changes that impact the reactions on the reactor vessel nozzles must meet the code criteria for the current code of record (or a later code reconciled to the code of record). System Nozzle No. Size El. Moments in Foot Pounds Forces in Pounds Mx My Mg Fx Fy Fg Main Steam Main Steam N3A N3B 24" 24" 310' 310' 12,480 -12,480 6,202 -6,202 -166,515 -166,515 -5,358 -5,358 1,813 1,813 -311 311 Steam Emerg. Cooling Steam Emerg. Cooling N5A N5B 10" 10" 305.8' 305.8' 6,790 -7,999 10,321 -8,288 11,143 12,876 2,335 2,323 -853 -993 -741 810 Feedwater Feedwater Feedwater Feedwater N4A N4B N4C N4D 10" 10" 10" 10" 295.9' 295.9' 295.9' 295.9' 76 401 -401 -76 7,163 8,071 -8,071 -7,163 -18,901 -22,954 -22,954 -18,901 -892 1,286 -1,286 -892 1,207 1,377 -1,377 1,207 582 -642 642 -582 Core Spray Core Spray N6A N6B 6" 6" 294.75' 294.75' 3,663 -8,105 4,776 -2,042 -8,982 243 -887 -1,273 1,537 4,213 518 -1,209 Recirculation Out PS-1 Recirculation Out PS-2 Recirculation Out PS-3 Recirculation Out PS-4 Recirculation Out PS-5 Recirculation In PD-6 Recirculation In PD-7 Recirculation In PD-8 Recirculation In PD-9 Recirculation In PD-10 N1A N1B N1C N1D N1E N2A N2B N2C N2D N2E 36"/28" " " " " 28" 28" 28" 28" 28" 274.75' " " " " - - - - - 1,957 -10,100 24,934 -29,921 -4,551 34,553 26,001 54,739 94,686 -22,903 19,551 -175 -3,011 -3,020 -16,089 1,989 4,434 -2,780 -2,788 43,550 49,641 -24 39,558 35,939 41,736 -41,114 -22,574 81,771 -26,790 -84,793 5,774 -753 669 -933 3,028 -1,035 -630 669 -933 -5,484 -1,017 -3,785 11,923 11,923 -2,656 3,824 -4,094 11,923 11,923 4,470 1,547 -76 1,198 1,007 -429 -610 572 1,198 1,007 5,610 Drain N18 2" 259.4' 94 -156 110 54 -28 NMP Unit 1 UFSAR Section XVI XVI-131 Rev. 25, October 2017 TABLE XVI-4 EFFECT OF VALUE OF INITIAL FAILURE PROBABILITY Strain to Initial Pf for Produce 4 Percent Strain Value of n 10Pf 0.5 x 10-7 2 8.6% 0.9 x 10-6 4 8.0% 1.0 x 10-5 6 10.0% NMP Unit 1 UFSAR Section XVI XVI-132 Rev. 25, October 2017 TABLE XVI-5 SINGLE TRANSIENT EVENT FOR REACTOR PRESSURE VESSEL 18-8 Stainless Steel Carbon Steel's and Condition Peak Strain Range Ni Cr - Iron Alloy Design Limit 4.0% 5.0% (ASME Section III) Safety Margin 8.0% 9.0% NMP Unit 1 UFSAR Section XVI XVI-133 Rev. 25, October 2017 TABLE XVI-6 POSTULATED EVENTS 1. Design Events (Expected During Vessel Lifetime) a. Normal heatup 100°F/h b. Normal cooldown 100°F/h c. Core spray initiation d. Feedwater flow interruption e. Blowdown for ten minutes plus normal 100°F/h cooldown f. Bolt-up cycle g. CRD coolant on-off h. Earthquake plus steady-state i. Unbolt head j. Design pressure 1250 psig k. Scram l. Jet thrust during pipe break 2. Nondesign Events (Not Expected During Vessel Lifetime) a. Cooldown 300°F/h b. Heatup 300°F/h c. Cooldown 600°F/h d. Heatup 600°F/h e. Overpressure 2500 psig f. Drowned vessel with cold water no insulation credit (550°F to 70°F) g. Overpressure 3750 psig (vessel destruction) NMP Unit 1 UFSAR Section XVI XVI-134 Rev. 25, October 2017 TABLE XVI-7 MAXIMUM STRAINS FROM POSTULATED EVENTS Events*** Peak Strain % 1. Stabilizer Bracket 1.h-1.1 0.12 2. Basin Seal Skirt 1.a-2.a 1.17 3. Control Rod Drive 1.h-1.g 0.58 4. Studs 1.a-1.i 1.03 5. Flanges 1.e-1.i 0.28 6. Steam Outlet Nozzle 1.j 0.19 7. Core Spray Nozzle 1.c 0.42 8. Recirculation Outlet 1.j 0.18 9. Recirculation Inlet 1.j 0.081 10. Feedwater Nozzle 1.d 0.338 11. Core Internals 1.a-1.e 0.340 12. Support Skirt 1.a-1.k- 2.a-1.e 0.52 13. Flanges 2.f 6.0 14. Basin Seal Skirt 2.c 2.34* 15. Several Discontinuities** 2.e 10* 16. Several Discontinuities** 2.g 40* 17. Basin Seal Skirt 2.d 2.34*

• Estimates only. ** Such as items 1-14 above. *** From Table 6.

NMP Unit 1 UFSAR Section XVI XVI-135 Rev. 25, October 2017 TABLE XVI-8 CORE STRUCTURE ANALYSIS RECIRCULATION LINE BREAK Structural Component Actual Applied Differential Pressure - psi Resultant Stress - psi Collapse Loading or ASME Code Allowable Loading - psi (See Notes Below) Safety Margin Lower Shroud and Core Support 125 (Inward) 7500 320* 2.6* Upper Shroud 7 (Outward) 420 260** 35** Core Support Plate 132 (Downward) - 250* 1.9* Guide Tube 132 (Outward) 4300 15,800*** 3.7** Fuel Channel 22 (Inward) - 30* 1.3* NMP Unit 1 UFSAR Section XVI XVI-136 Rev. 25, October 2017

• Loading, or margin, based on collapse stress. ** Loading, or margin, based on ASME Code allowable stress. *** ASME Code allowable stress.

NMP Unit 1 UFSAR Section XVI XVI-137 Rev. 25, October 2017 TABLE XVI-9 CORE STRUCTURE ANALYSIS STEAM LINE BREAK Structural Component Actual Applied Differential Pressure - psi Resultant Stress Due to Break Inside Flow Limiter - psi Collapse Loading or ASME Code Allowable Loading - psi (See Notes Below) Safety Margin Break Outside Flow Limiter Break Inside Flow Limiter Lower Shroud and Core Support 40 (Outward) 63 (Outward) 3800 15,800*** 4.2** Upper Shroud 13 (Outward) 22 (Outward)**** 1400 15,800*** 11.3** Core Support Plate 27 (Upward) 41 (Upward) 7000 15,800*** 2.3** Guide Tube 27 (Inward) 41 (Inward) 1400 106* 2.6* Fuel Channel 14 (Outward) 23 (Outward) - 30* 1.3*

• Loading, or margin, based on collapse stress. ** Loading, or margin, based on ASME Code allowable stress. *** ASME Code allowable stress. **** A pressure drop margin for the upper shroud differential pressure of 22 psi exists for operation outside of the RIP region specified on the power flow operating maps. However, normal operation within the RIP region may exceed 22 psi in the event of a MSLB. Plant operating procedures restrict operation in the RIP region.

NMP Unit 1 UFSAR Section XVI XVI-138 Rev. 25, October 2017 TABLE XVI-9a CORE SHROUD REPAIR DESIGN SUPPORTING DOCUMENTATION Document Number Description GENE-B13-01739-04 NMP1 Shroud and Shroud Repair (NMPC Calculation Hardware Analysis #SQ-Vessel-M028) GENE-B13-01739-05 Safety Evaluation for Installation (NMPC 50.59 Evaluation of Stabilizers on the NMP1 Core 94-080) Shroud GENE-B13-01739-03 Seismic Design Report of the Shroud (NMPC Calculation Repair for NMP1 Power Plant #SQ-Vessel-M027) 24A56426 Stress Report, "Shroud & Stabilizers (NMPC Calculation Code Design Specification - Shroud #SQ-Vessel-M026) Stabilizers" 25A5583 Design Specification, "Shroud Repair Hardware" 107E5679, Sheets 1-4 Modifications & Installation Drawings 25A5584 Fabrication Specification, "Fabrication of Shroud Stabilizer" FDI 0245-90800 Field Disposition Instruction 25A5585 Installation Specification, "Stabilizer Installation" 21A2040 Cleaning and Cleanliness Control 24A5586 Shroud Stabilizer Code, Design Specification GENE-771-44-0894 Justification of Allowable Displacements of the Core Plate and Top Guide - Shroud Repair NMP Unit 1 UFSAR Section XVI XVI-139 Rev. 25, October 2017 TABLE XVI-9a (Cont'd.) Document Number Description GENE-B13-01739-5.1 Modification to GE Core Shroud (NMPC 50.59 Evaluation Repair Design 96-018) NMPC 50.59 Evaluation Core Shroud Vertical Weld Repair 98-103 Clamp NER-IM-059 NMP-1 Core Shroud Vertical Weld Repair Design Report, MPR Report No. MPR-1966 NRC Safety Evaluation NMP1 Core Shroud Repair, dated 3/31/95 NRC Safety Evaluation Modifications to Correct Core Shroud Repair Deviations, dated 3/3/97 NRC Safety Evaluation Modifications to Core Shroud Stabilizer Lower Wedge Retaining Clip and Evaluation of Shroud Vertical Weld Cracking, dated 5/8/97 NRC Safety Evaluation Modification of Core Shroud Tie Rod Upper Spring Assemblies, dated 6/7/99 NRC Safety Evaluation Modification of Core Shroud Tie Rod Upper Support and Tie Rod Nut Assemblies, dated 10/3/07 NMP Unit 1 UFSAR Section XVI XVI-140 Rev. 25, October 2017 TABLE XVI-10 DRYWELL JET AND MISSILE HAZARD ANALYSIS DATA Break Case Missile Travel Distance (ft) Jet Area at Break Line Opening (ft2) Total Jet Force (lb) Missile Min. Prob. Impact Area Impact Energy (ft-lb) Main Steam Line - Vertical Break of 90° Elbow Outlet of Reactor A 1 2.84 124,000 24-in pipe (5,050 lb) Line 10 ft long, 0.5 ft wide 124,000 Recirculating Line - Vertical Break at 90° Elbow - El 274'-9" B 8 4.00 249,000 Normal to Shell 28-in pipe Point on 90° elbow 2,590,000 Feedwater Line - Vertical Break at 90° Ell on Reactor C 4 0.394 37,800 10-in pipe (1,620 lb) Line 14 ft long, 0.25 ft wide 151,200 Safety Valve - Horizontal Break at Inlet to Valve D 4 0.31 18,400 Valve (850 lb) 7 in2 68,500 Recirculating Line - Horizontal Break at Discharge Valve - Swings in Line with Baffle E - 4.00 239,000 Normal to Shell None None None Reactor Recirculating Pump Valve F 17 1.22 102,000 Valve Bonnet (8,590 lb) 50.3 in2 2,000,000 NMP Unit 1 UFSAR Section XVI XVI-141 Rev. 25, October 2017 TABLE XVI-11 DRYWELL JET AND MISSILE HAZARD ANALYSIS RESULTS Case Total Energy to be Absorbed (ft-lb) Energy Absorbed (ft-lb) Force to Shear (lb) Force (lb) Deflection (in) Results A 149,940 154,900 19,656,000 46,600 2.5 No Rupture B 2,750,000 2,770,000 10,000,000 4,000,000 8.2 No Rupture C 162,225 167,190 26,676,000 66,405 3.5 No Rupture D 75,550 75,600 428,840 315,000 4.6 No Rupture E -- -- 5,513,300 239,000 0.13 No Rupture F 2,080,000 2,080,000 1,960,000 1,400,000 9.5 No Rupture NMP Unit 1 UFSAR Section XVI XVI-142 Rev. 25, October 2017 TABLE XVI-12 STRESS DUE TO DRYWELL FLOODING INDICATED DRYWELL STRESSES AT BASE OF SAND CUSHION (EL 222'-8 1/4") Loading Considered No Earthquake With Earthquake Allowable(2) Meridian Circumferential Meridian Circumferential Tensile Comprehensive Flooded to El 301' + E.Q. -5,250 12,600 -9,960 +8,370 +17,300 +19,250 -15,500(3) Flooded to El 333' + E.Q. -4,030 +14,600 -8,730 +9,900 +18,850(1) +19,250 -16,000(3) NOTES: (-) Indicates compression. (+) Indicates tension. (1) Maximum stress--does not occur during maximum compressive stress. (2) Allowable stress with no increase for earthquake condition. (3) Increased buckling stress due to internal pressure.

Reference:

Y. C. Gung and E. E. Sechler, "Buckling of Thin-Walled Circular Cylinders Under Axial Compression and Internal Pressure," and L. A. Harris, et al, "The Stability of Thin-Walled Unstiffened Circular Cylinders Under Axial Compression Including the Effects of Internal Pressure," and CB&I review of the above reports. NMP Unit 1 UFSAR Section XVI XVI-143 Rev. 25, October 2017 TABLE XVI-13 ALLOWABLE WELD SHEAR STRESS Normal Allowable Shear Stress (AISC 1969) 21,000 psi Allowable Shear Stress - Increased 50 Percent 31,500 psi NMP Unit 1 UFSAR Section XVI XVI-144 Rev. 25, October 2017 TABLE XVI-14 LEAK RATE TEST RESULTS Vessel Tested Nominal Pressure psig Vessel Condition Time Period Each Day Leak Rate (% per day) Drywell Supp. Chamber Drywell Supp. Chamber Uncorrected Temp Corrected Temp & Hum. Corrected Temp, Hum. & Reference Chamber Humidity Corrected Drywell 62 0 Dry Dry 3:00 am 6:00 am 0.027 0.038 (0.020)* 0.043 Drywell 37 35 Wet Wet 5:00 am 7:00 am 0.027 0.024 0.050 0.050 Supp. Chamber 39 35 Dry Dry 3:00 am 6:00 am 0.024 0.025 0.017 0.024 Supp. Chamber 37 35 Wet Wet 4:00 am 6:00 am 0.019 0.018 0.027 0.027 Both 35 35 Wet Wet 12:00 am 6:00 am 8:00 pm 10:00 pm 0.012 0.016 0.014 0.017 (0.006)* 0.009 0.008 0.009

• Gain Shown by ( ).

NMP Unit 1 UFSAR Section XVI XVI-145 Rev. 25, October 2017 TABLE XVI-15 OVERPRESSURE TEST--PLATE STRESSES Percent Percent Code Plate Area Allowable Stress 100 71 78 84 65 95 36 102 17 105 NMP Unit 1 UFSAR Section XVI XVI-146 Rev. 25, October 2017 TABLE XVI-16 STRESS SUMMARY Total Stress (Thermal + Primary + Seismic) - psi Static Analysis Dynamic Analysis Containment Spray System A Inside Drywell - Elevation 306 Elevation 284 Elevation 255 Elevation 245 Outside Drywell - Pump to Heat Exchanger Heat Exchanger to Drywell Drywell to Suppression Chamber Inside Suppression Chamber 15,890 19,103 12,187 13,866 17,840 17,012 21,173 21,449 19,319 13,970 10,216 12,248 13,877 18,340 17,913 13,797 23,419 17,519 Containment Spray System B Inside Drywell - Elevation 279 Elevation 260-6 Outside Drywell - Pump to Heat Exchanger Heat Exchanger to Drywell Test Lines to Suppression Chamber 13,900 11,396 11,299 12,344 18,219 15,400 13,331 12,898 13,692 19,963 Core Spray Outside Drywell - Pump to Topping Pumps Topping Pump to Drywell Condensate Makeup Pump Cross Connection Test Line Inside Drywell - East West 8,456 21,140 11,491 12,053 9,156 27,150 25,750 NMP Unit 1 UFSAR Section XVI XVI-147 Rev. 25, October 2017 TABLE XVI-17 HEAT TRANSFER COEFFICIENTS AS A FUNCTION OF DROP DIAMETER Overall Heat Transfer Drop Diameter (d) Coefficient (H) (microns) (Btu)/(ft2)(hr)(F) 100 11,600 300 3,860 500 2,320 1,000 1,160 NMP Unit 1 UFSAR Section XVI XVI-148 Rev. 25, October 2017 TABLE XVI-18 HEAT TRANSFER COEFFICIENT AS A FUNCTION OF PRESSURE Surface Heat Transfer Pressure Coefficient (Us) (psia) (Btu)/(ft2)(hr)(F) 14.7 49,600 40.0 104,500 50.0 122,100 NMP Unit 1 UFSAR Section XVI XVI-149 Rev. 25, October 2017 TABLE XVI-19 RELATIONSHIP BETWEEN PARTICLE SIZE AND TYPE OF SPRAY PATTERN Comparative Type Particle Size Range of Spray in (Microns) Particle Size Range 2 - 5 Dry Fog 10 - 40 Wet Fog 50 - 100 Misty Rain 200 - 400 Light Rain 500 - 1000 Moderate Rain 2000 - 5000 Heavy Rain NMP Unit 1 UFSAR Section XVI XVI-150 Rev. 25, October 2017 TABLE XVI-20 ALLOWABLE STRESSES* FOR FLOOR SLABS, BEAMS, COLUMNS, WALLS, FOUNDATIONS, ETC. 1. Reactor Building 2. Waste Disposal Building 3. Screen and Pump House Dead Load Plus Live Load Plus Operating Load Stresses Considered Plus Design Earthquake Reinforcing steel--maximum allowable tensile stress 0.5 Fy = 20,000 psi Reinforcing steel--maximum allowable compressive stress 0.34 Fy = 13,600 psi Concrete--maximum allowable compressive stress 0.45 F'c = 1,575 psi Concrete--maximum allowable shear stress 1.1 = 65 psi Reinforced concrete shear walls--allowable unit stress 2.67 = 160 psi Concrete--maximum allowable peripheral shear stress 2 = 118.8 psi Concrete--maximum allowable bearing stress 0.25 F'c (Full A) = 875 psi 0.375 F'c (1/3 A) = 1,312 psi NOTES: 1. Rebar Steel--Fy = 40,000 psi. 2. Concrete--F'c = 3,500 psi. 3. 11% ground acceleration. 4. 5% damping.

• Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section XVI XVI-151 Rev. 25, October 2017 TABLE XVI-21 ALLOWABLE STRESSES* FOR STRUCTURAL STEEL 1. Reactor Building (Superstructure, Columns, Beams, Roof Steel, Bracing, Crane Supports) 2. Waste Disposal Building (Same Components as Reactor Building) 3. Screen and Pump House (Same Components as Reactor Building) 4. Drywell Radial Steel Framing (Beams Only) Stresses Considered Dead Load Plus Live Load Plus Operating Load Plus Design Earthquake Dead Load Plus Live Load Plus Operating Load Plus Wind Tension on net section 0.6 Fy = 21,600 psi 0.8 Fy = 29,000 psi Shear on gross section 0.4 Fy = 14,400 psi 0.533 Fy = 19,200 psi Bending tension and compression 1.0 1.0 A325 H.S. bolts: Tension Shear 0.5 Fy = 40,000 psi 0.19 Fy = 15,000 psi 0.667 Fy = 53,300 psi 0.25 Fy = 20,000 psi NOTES: 1. 11% ground acceleration. 2. 5% damping.

• Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section XVI XVI-152 Rev. 25, October 2017 TABLE XVI-22 ALLOWABLE STRESSES* - REACTOR VESSEL CONCRETE PEDESTAL Stresses Considered Dead Load Plus Equipment Load Plus Temperature (Operating) Dead Load Plus Equip. Load Plus Jet Load Plus Temp. Plus Design Earthquake Reinforcing steel--maximum allowable tensile stress 0.25 Fy = 10,000 psi 0.50 Fy = 20,000 psi Reinforcing steel--maximum allowable compressive stress 0.05 Fy = 2,000 psi 0.10 Fy = 4,000 psi Concrete--maximum allowable compressive stress 0.133 F'c (Bending) = 533 psi 0.116 F'c (Direct) = 465 psi 0.133 F'c (Bending) = 533 psi 0.116 F'c (Direct) = 465 psi Concrete--maximum allowable shear stress 0.55 = 35 psi 1.1 = 70 psi NOTES: 1. Jet plus seismic loads per J. A. Blume's "Earthquake Analysis: Reactor Pressure Vessel." 2. Temperature = 60°F maximum gradient. 3. Concrete F'c = 4,000 psi. 4. 5% damping. 5. 11% ground acceleration. 6. Reactor building deflection at stabilizers = 72 mils. 7. Stresses summarized in XVI-A (Figure XVI-9) show allowable stresses one-third over normal Code values. All above-normal stresses are for localized areas except at "D" (anchor bolts), which show the general stress condition.

• Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section XVI XVI-153 Rev. 25, October 2017 TABLE XVI-23 DRYWELL - ANALYZED DESIGN LOAD COMBINATIONS Load Condition No. 1 2 3 4 5 6 7 8 9 10 1. External pressure x x x x 2. Internal pressure x x x x 3. Vertical live loads x x x x x x x x x x 4. Earthquake force (horiz.) x x x x x 5. Earthquake force (vert.) x x x x x 6. Lateral earthquake reactions from internal structure x x x x x 7. Water load during refueling x x 8. Deadweight of vessel x x x x x x x x x x 9. Lateral jet reactions from internal piping x x x x 10. Flooded to el 333-0 x x NOTES: 1. 30% horizontal earthquake acceleration. 2. 5.5% vertical earthquake acceleration. 3. 2.5% damping. 4. All stresses are within normal Code* allowables.

• Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section XVI XVI-154 Rev. 25, October 2017 TABLE XVI-24 SUPPRESSION CHAMBER - ANALYZED DESIGN LOAD COMBINATIONS Loading Condition No. 1 2 Internal pressure - 35 psi x Water load: (a) 94,000 cu ft (incident) x (b) Full (postincident) x Flooding to el 301-0 (in drywell) x Horizontal earthquake acceleration x x Vertical earthquake acceleration x x Thermal x x NOTES: 1. 15% horizontal earthquake acceleration. 2. 5.5% vertical earthquake acceleration. 3. 2.5% damping. 4. Stresses within normal Code allowables except one-third increase in suppression chamber support columns for Loading Condition 2. NMP Unit 1 UFSAR Section XVI XVI-155 Rev. 25, October 2017 TABLE XVI-25 ACI CODE 505 ALLOWABLE STRESSES AND ACTUAL STRESSES FOR CONCRETE VENTILATION STACK Loading Condition Reinforcing Steel Tensile Stress Concrete Compressive Stress Concrete Shear Stress Allowable Actual Allowable Actual Allowable Actual Dead load + wind load 0.375 Fy = 15,000 psi 11,500 psi 0.375 F'c = 1,500 psi 936 psi 1.1 = 70 psi 14 psi Dead load + design earthquake load (ACI Code 505) 0.375 Fy = 15,000 psi 14,850 psi 0.375 F'c = 1,500 psi 1,260 psi 1.1 = 70 psi 25 psi Dead load + design earthquake load (Dynamic analysis) 0.375 Fy = 15,000 psi 36,000 psi 0.375 F'c = 1,500 psi 1,580 psi 1.1 = 70 psi 44 psi NOTES: 1. Fy minimum based on mill test reports = 45,000 psi. 2. Maximum wind velocity = 110 mph. 3. Maximum temperature differential = 100°F. 4. 11% horizontal ground acceleration. 5. 7.5% damping. NMP Unit 1 UFSAR Section XVI XVI-156 Rev. 25, October 2017 TABLE XVI-26 ALLOWABLE STRESSES* FOR CONCRETE SLABS, WALLS, BEAMS, STRUCTURAL STEEL, AND CONCRETE BLOCK WALLS 1. Turbine Building - Class II 2. Control Room - el 277 Class I 3. Battery Room - el 277 Class I 4. Auxiliary Control Room - el 261 Class I 5. Battery Board Room - el 261 Class I 6. Diesel Generator Area - el 261 Class I Dead Load Plus Live Load Plus Operating Load Stresses Considered Plus Design Earthquake Reinforcing steel--allowable tensile stress 0.5 Fy = 20,000 psi Reinforcing steel--allowable compressive stress 0.34 Fy = 13,600 psi Concrete--allowable compressive stress 0.45 F'c = 1,575 psi Concrete--allowable shear stress 1.1 = 65 psi Reinforced concrete shear walls-- allowable shear stress 2.67 = 160 psi Structural steel--allowable bending stress 36 ksi material 0.6 Fy = 21,600 psi 50 ksi material 0.6 Fy = 30,000 psi Structural steel--allowable web shear stress 36 ksi material 0.4 Fy = 14,500 psi 50 ksi material 0.4 Fy = 20,000 psi Reinforced concrete block walls-- allowable mortar unit stress 0.04 F'c = 36 psi NMP Unit 1 UFSAR Section XVI XVI-157 Rev. 25, October 2017 TABLE XVI- NOTES: 1. 11% horizontal ground acceleration. 2. 5% damping.

• Also see Section XVI, Subsection G.

NMP Unit 1 UFSAR Section XVI XVI-158 Rev. 25, October 2017 TABLE XVI-27 SYSTEM LOAD COMBINATIONS System Load Combinations No. I II III 1. Main steam 01 Yes Yes Yes 2. Control rod drive Control rod drive Control rod drive 28 & 44 44.1 53 Yes Yes Yes No Yes Yes No Yes Yes 3. High-pressure reactor feedwater High-pressure reactor feedwater High-pressure reactor feedwater 29 30 31 Yes Yes Yes N/A N/A Yes Yes Yes Yes 4. Reactor recirculation 32 Yes No No 5. Reactor cleanup 33 Yes Yes Yes 6. Reactor head spray 34 Yes Yes Yes 7. Reactor shutdown cooling (partial) Reactor shutdown cooling (partial) 38 38 Yes Yes Yes N/A Yes N/A 8. Emergency cooling 39 Yes Yes Yes 9. Reactor core spray Reactor core spray 40 81 Yes Yes Yes Yes Yes Yes 10. Liquid poison (partial) Liquid poison (partial) Liquid poison (partial) 41 42 42.1 Yes Yes Yes N/A N/A Yes N/A N/A N/A 11. Radioactive waste disposal (partial) Radioactive waste disposal (partial) 45 45 Yes Yes N/A N/A N/A No 12. Condensate pump inlet and discharge Condensate pump inlet and discharge 49 50 Yes Yes N/A N/A No Yes 13. Reactor feedwater booster pump discharge 51 Yes N/A Yes 14. Fuel pool cooling and filtration 54 Yes N/A Yes 15. Condensate storage and transfer 57 Yes N/A No 16. Drywell and suppression chamber vacuum relief 68 Yes Yes N/A NMP Unit 1 UFSAR Section XVI XVI-159 Rev. 25, October 2017 TABLE XVI-27 (Cont'd.) System Load Combinations No. I II III 17. Reactor building closed loop cooling 70 Yes N/A Yes 18. Service water and/or emergency service water 72 Yes N/A Yes 19. Diesel generator cooling water 79 Yes N/A Yes 20. Containment spray (piping and manual valves) Containment spray (all other components) 80 80 Yes Yes Yes Yes Yes Yes 21. Diesel generator fuel oil 82 Yes N/A No 22. Containment spray raw water cooling 93 Yes N/A No 23. Diesel generator start air system 96 No N/A N/A 24. Fire protection (water) 100 Yes N/A No 25. Fire protection (transformer) 101 Yes N/A No 26. High-pressure instr. piping outside drywell High-pressure instr. piping inside drywell and low-low water level and drywell pressure 108 108 Yes Yes N/A Yes N/A Yes 27. Condenser offgas Condenser offgas Condenser offgas 76 77 112 Yes Yes Yes N/A N/A N/A N/A N/A N/A 28. Instrument air 94 & 113 Yes N/A N/A 29. Breathing air 114 Yes N/A N/A 30. Emergency ventilation Emergency ventilation 202 201.2 Yes Yes N/A N/A N/A N/A NOTES: I = Normal and seismic. II = Normal, seismic, and accident. III = Normal and water hammer. Yes = Load combination considered and accounted for in design. No = Load combination not considered or accounted for in design. N/A = Load combination not applicable. NMP Unit 1 UFSAR Section XVI XVI-160 Rev. 25, October 2017 TABLE XVI-28 HIGH-ENERGY SYSTEMS Inside Containment Main steam Feedwater Reactor recirculation Core spray Containment spray Emergency condenser supply and return Control rod drive hydraulic Liquid poison Relief valve discharge Shutdown cooling Head spray Cleanup NMP Unit 1 UFSAR Section XVI XVI-161 Rev. 25, October 2017 TABLE XVI-29 HIGH-ENERGY SYSTEMS Outside Containment Main steam (Class II) Feedwater (Class I) Core spray (Class I) Containment spray, including raw water (Class I) Control rod drive (Class I) Liquid poison (Class I) Emergency condensers (Class I) Reactor cleanup (Class I) Reactor shutdown cooling (Class I) Reactor head spray (Class I) Electric steam boiler (Class I) Carbon dioxide fire protection (Class I) NMP Unit 1 UFSAR Section XVI XVI-162 Rev. 25, October 2017 TABLE XVI-30 SYSTEMS WHICH MAY BE AFFECTED BY PIPE WHIP Safeguard Systems Containment spray Core spray Liquid poison Emergency condensers High-pressure coolant injection Control rod drive Diesel generators Shutdown Systems Instrument air Service water Reactor building closed loop cooling Spent fuel pool cooling Shutdown coolant Nonshutdown Systems Fire protection Condensate storage and transfer Cleanup Electric steam boiler NMP Unit 1 UFSAR Section XVI XVI-163 Rev. 25, October 2017 TABLE XVI-31 CAPABILITY TO RESIST WIND PRESSURE AND WIND VELOCITY External Pressure Internal Pressure Siding Structural Steel Framing Access Openings (Doors) Design Value (psf) Blowout Panels (psf) Roof (psf) Wall (Siding) (psf) (psf) (mph) (psf) (mph) Superstructures Reactor Building East and West Walls North Wall - 150 140 - 250 235 - 90 90 - 190 190 40(1) - - 65 - - 144 - - - 150 140 Turbine Building Battery Room Battery Board Room Control Room Aux. Control Room Diesel Gen. Area Diesel Gen. Board Room 165 210 210 300 300 190 230 250 285 285 340 340 270 300 90 N/A N/A 140 140 135 230 190 N/A N/A 235(2) 235(2) 230 300 40 40(1) 40(1) - - 40 - 62 N/A N/A N/A N/A N/A N/A 144 N/A N/A N/A N/A N/A N/A 165 72 72 300 300 190 230 Screen House 60 150 60 150 40(1) N/A 144 60 Substructures Reactor Building (Below el 340) 230 300 - - 40(1) - 430 430 Screen House (Below el 261) 230 300 - - - - 430 430 Waste Building (Below el 261) 230 300 - - - - 430 430 Turbine Building (Below el 261) 230 300 - - - - 430 430 NMP Unit 1 UFSAR Section XVI XVI-164 Rev. 25, October 2017 TABLE XVI-31 (Cont'd.) External Pressure Internal Pressure Siding Structural Steel Framing Access Openings (Doors) Design Value (psf) Blowout Panels (psf) Roof (psf) Wall (Siding) (psf) (psf) (mph) (psf) (mph) Other Ventilation Stack 63(3) 177(4) - - - - - - - Ventilation Stack 98(3) 222(5) - - - - - - (1) One-third increase in stresses* above normal. (2) Modification of the structural support steel is in process so that it will be capable of withstanding 300 mph wind pressure (230 psf). (3) Includes 0.6 shape factor and 1.3 gust factor. (4) Wind velocity at the top of the stack; corresponds to a 145-mph design wind 30 ft above the ground with stress levels at yield. (5) Wind velocity at the top of the stack; corresponds to a 175-mph design wind 30 ft above the ground with stress levels approaching failure.

• Also see Section XVI, Subsection G.

.... . . ' . . " '*. :* b. 17 . .. 4 f. ) SEISMIC ANALYSIS OF REACTOR VESSEL GEOMETRIC AND LUMPED MASS REPRESENTATION EL.3241-7 9/3211 +-(0 "' O') EL.303'-g" ...J v UJ (/) EL.2891-§ II (/) +-UJ > 0:: 0 -(\j EL.281!..6" <<) + <.) . <( 'It UJ 0:: +-EL .2.7 41-I O . <D v EL.2641-811 I() -.. p A.* . I " A l.D. I -4" :.*A ...J <D <( Cl 011 " <D (/) 1£.1 -0 E'L.242.'-6" UJ + a. a:: 0 u <( UJ -<D .. Q:'. en .j) u) .:*A' A':A -,._ C\I (\J ....J ....J <( 3= Q ....J Ll.J ::c CJ) ,._ N N (K*SEC.2\ POINT MASS FT. J l I 33,8 2 39.8 3 21.0 4 30.4 5 18.2 6 20.4 7 10.4 GEOMETRIC FIGURE LUMPED MASS SYSTEM FIGURE XVl-1 UP&Alt Rev. 14 1991) ( 1-UJ UJ LL.. :z: S2 ..... < > UJ ...J UJ REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. MOMENT 310 ""' ..... h \ LIZER STR. EL 303' -9" 280 I ' 300 * % -__, _ ___,_--+---+---+----+----+----+----+---+---+---+---1 I ' f2> I % j_ 290 II ' ' \ \?\ I l \' \ \ \\\ *"*---+---+---+---+---+---+--+---+---+--+--+--+----I 210 H _ ':e ;:". If! 'l,' \ i"\ 260 \5::+::1=3 BASE -EL. 259' -5"--+----+---+---+---+---+----1 \I\. 250 ' 240 l ""-* SEISMIC-(INCLUDES EFFECT OF BLDG. \ MOVEMENT AND BASE ' "" \A_. JET THRUST """ 230 \ ' iO,oo \ ;tt . .rQ \ o41c 1 l"'iii... BASE EL. 225' -6" I -........ 0 10 20 30 40 50 60 MOMENT (FT.-KIPS) X 103 FIGURE XVl-2 UFSAR Rev. 14 (June 1981) ( 310 300 290 280 I-u.J u.J u... z: 0 I-270 < > u.J ...J u.J 260 250 240 230 REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. SHEAR -----* " I I I ' I J I I I I I . I I < I i\ I J" I I ; I I\. ;r ----ff} I'\ I\. I *t I '\ I I 111.. \ I I I \ I I JET THRUST , ti' I '(;, I I I I I I ' I I I \ I 0 200 400 600 800 1000 SHEAR* KIPS FIGURE XVl-3 UFSAR Rev. 14 (June 1996) 310 300 290 280 I-UJ UJ u.. z: 270 0 < > UJ ....J UJ 260 250 240 230 ....J ....J -== ._;;;= Cl ....J !::!::! REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. DEFLECTION I / J /, v !"--.JET THRUST COND./ ' I -SEISMIC I v I " J § fj /v fl-0 '--..J v / , // / J I l v INCLUDES THE EFFECT OF BUILDING MOVEMENT I ,Y I' ,-7 -I .. c.3 CJ ,_ a: 0... ii? 0 100 200 DEFLECTION (MILLS) RGURE XVl-4 UFSAR Rev. 14 (June 1991) ( \ 310 300 290 280 w w I z 0 270 w ...J LLJ 260 250 240 230 REACTOR SUPPORT DYNAMIC ANALYSIS ELEVATION VS. ACCELERATION INCLUDES THE EFFECT OF 1 I BLDG. MOVEMENT a BASE ROCKING :i "' < 3: a ..J "' , 15 / , / j I 1Y ! / I 7-o/ o/ _}... _;/ ..:::> ">) I I BASE EL. 2251-6 10 20 30 40 50 60 ACCELERATION -PERCENT GRAVtTY FIGURE XVl-5 UFSAR Rev. 14 (.June 19H) Nine Mile Point Unit 1 FSAR FIGURE XVI-6 THRU FIGURE XVI-8 FIGURES XVI-6 THRU XVI-8 HAVE BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 REACTOR VESSEL SUPPORT STRUCTURE STRESS SUMMARY LOADINGS STRESSES CPS I) OPERATING + OPERATINGa OPERATING + EARTHQUAKE + EARTHQUAKEa JET THRUST POINT OF ALLOWABLE ACTUAL ALLOWABLE ACTUAL ALLOWABLE ACTUAL DESCRIPTION STRESS STRESS STRESS STRESS STREssb STRESS CONCRETE -Bear Ing 1250 435 1250 1230 1666 1660 Alm CONCRETE FR I CT I ON -t-. Bushharrrnered Surface 100 45" 100 4511 130 4511 A2c . CONCRETE KEY -Bear Ing 1000 1000 725 1333 1201) CONCRETE SHEAR PLANE 140 45 140 137 185 i-15 Cl RE I NFORC I NG -Tens Ion 20000 9100d 20000 15500d 26600 18000d C2 CONCRETE -Compress I on 2250 585j 2250 1380j 3000 1840j v ANOiOR BOLTS -Tension 21600 21600 15800 48500i 31300 v ANCHOR BOLTS -Shear 16200 o" 16200 om 21600 om ANCHOR BOLTS -Tension 21600 21600 8000 28000 12200 CONCRETE -Bearing 1875 94 1875 294 2500 450 REINFORCING -Tension 20000 4900e. 20000 14500e. 26600 18500e. CONCRETE -Compress Ion 2250 2250 132 3000 186 RE I NFORC I NG -Compression 16000 0 16000 930 -21300 1300 REINFORCll{; -Tension 20000 1206 20000 147006 26600 209006 REINFORCING .:. Compression 16000 1140h 16000 2900h 21300 3260h CONCRETE -Bearing 875 120 875 390 1170 455 CONCRETE KEY -Bearing 875 875 770 1170 1040 A-490 BOLT -Tens I on 50000 PRESTRESS 3460oP 29000 7430oB 68000 A-490 BOLT -Shear 24000 2430! 24000 9630! 26600 16050! CONCRETE -Compression 2250 ask 2250 217k 3000 271k REINFORCING -Tension 20000 20000 9700 26600 13700 CONCRETE SHEAR PLANE 140 26 140 80 185 100 Code allowable stresses 150 ps I therma I stress Inc I uded 1 /3 over norma I code stresses 85 ps I therma I stress Inc I uded -Dynamic loads only 2430 ps I therma I stress included 4300 ps I therma I stress inc I uded m Shear transmitted to key (see point A2) 4900 ps I therma I stress Inc I uded n 45 ps I therma I stress Inc I uded Includes 720 psi thermal stress Allowable stress= 50,000 -1.6 fv Allowable stress = 0.9 x functional fe 11 ure = 100,000 -1.6 'fv (fv = shear stress} C fv = shear stress l h 200 ps I therma I stress inc I uded .i Fy = yield stress = 54,000 psi CC-10351 0.9 x functional fal lure= 48,500 psi SH I ELD WALL CONCRETE SURFACE Al SHEAR KEY A2 w tL a:: u z 0 u 0 en 0:: 0 I u z <I: REACTOR VESSEL SUPPORT SKIRT REACTOR VESSEL BASE M EL.2601-4 1/4 PLATE A / EL.2571-10 1/2 CONCRETE r, REINFORCING G (HORI z.) I RE I N FORC I NG H (VERT.) ANillll E CONCRETE REACTOR SUPPORT STRUCTURE ( 5000 p s i CONC * ) 5 '-0 81-0 RADfUC: CONCRETE M (HOR I Z) RE I NFORC I NG M H.2531-9 1/2 c c EL. 249 I -6 CONCRETE SHEAR PLANE N 9'-0 RADIUS SHEAR KEY K :-. '. * -:*t> EL. 225 1 -6 _ * .. 4 -**SURAACE° J ::_..'_.: ... '. .. ei* .. I ., .: I **-* ** ,

• I I I BASE CONCRETE (3500 psi CONCRETE) FIGURE XVl-9 UFSAR Rev. 14 (June 1996)

( THERMAL ANALYSIS 115F SHIELD WALL VENTILATION : SLEEVE 114F -= .-.) VENTILATION SLEEVE --*. ;-_, !, -REACTOR VESSEL '., SUPPORT STRUCTURE / .. -, * .... VENTILATION OPENING / <i, ' ... INSULATION 125F TO 111 F TO 115F REACTOR VESSEL 546F CONTROL ROD DRIVES 111F TO 115F. FIGURE XVl-10 UFSAR Rev. 14 (June 1991) ( ' FAILURE PROBABILITY DENSITY FUNCTION 0 10 20 30 40 STRAIN, PERCENT FIGURE XV1-11 UFSAR Rev. 14 (June 19H} ADDITION STRAINS PAST 4% REQUIRED TO EXCEED DEFINED SAFETY MARGIN 8 6 ADDITIONAL STRAIN, 4 PERCENT 2 0 10*" o, ' ' LOW CYCLE FATIGUE EXTRAPOLATION ', __ o o--INCREASE OF 10 FROM -o INITIAL FAILURE PROBABILITY INITIAL FAILURE PROBABILITY FIGURE XVl-12 UflAR Rev. 14 (.June 1991) UPPER RING CENTRAL RING CENTRAL UPPER CYLINDER CENTRAL MID CYLINDER CENTRAL LOWER CYLINDER SOURCE: 105E1 413A SH. 1 < G. E. DWG, l XV I_ t 2o. dgn < coddl SHROUD WELDS HI ,,.---H2 H3 H4 HS ,,.----H6A H6B SHROUD SUPPORT RING '----INCONEL SHROUD SUPPORT SKIRT FIGURE XVl-120 UFSAR Rev. 14 June 1996 CORE SHROUD STABILIZERS TOP SPRING u......i.m--u LIMIT STOP TIE ROD CORE PLATE SPACER BOTTOM SPRING ll SOURCE1 SH.2 <G.E. DWG,> XVI-l2b.dgn <codd> SECTION L-L FIGURE XVl-12b UFSAR Rev. 14 June 1996 + 6.0f 31 * .25 H1 I + . ----H ,_.,vs 2.0 __ 18.50 t : 14 I 0 OI 90.12 0 01 H 5 lvs 1 6.3.50 j H6 A V13 4.SQ

• H6 8 22.1.3 t H ' CORE SHROUD WELDS I V3 V4 vs I I V7 V8 I IO 0 REPAIR CL.AMP !o 01 (TYP) V10 V11 rv12 . V14 I Jvis r,6 i---:---.._ SHROUD HEAD FUN CE TOP GUIDE SUPPORT CORE PLATE SUPPORT SHROUD SUPPORT RING FIGURE XVl-12c UFSAR Revision 16 November" 1qqq Left Bayonet Eccentric Threaded Pin vq1v10 VERTICAL WELD CLAMP ASSEMBLY """" Eccentric V9/V10 Vertical Weld Locking Screw V9/V1 o Plate FIGURE XVl-12d UFSAR Revision 16 November 1qqq lt r: G> :II i * "i"' ...& ...& -l ...& I -l) (/) 0. 70 60 so 40 ,,_ -......... 30 20 10 v / I 1 / / Initial Pressure = OPSIC Initial Water Temperature = 90 F v Dow ncomer Submergence = 4 ft ;' I'-l-1 v v / ( Drywell ' ' ... ""' __.,,,,.,. -Supression Chamber 10 Time After Accident (sec)

Nine Mile Point Unit 1 FSAR FIGURE XVI-14 THIS FIGURE HAS BEEN DELETED UFSAR Revision 14 1 of 1 June 1996 DRYWELL TO CONCRETE AIR GAP ........ , FORM CONSTRUCTION TYPICAL VENT PIPE DETAIL COMPRESSIBLE POLYURETHANE TAPE (1/2" x 3/4'.'* AL AT ALL FLANGE JOINTS) '. x o*-2n M. BOLT TYPICAL JOINT DETAIL .613 *871_J 2.438 .434..J .__moo!'"-1.33 .526_Jr'---+-DRYWELL GROWTH DUE TO PRESSURE AND TEMPERATURE (INCIDENT) FIGURE XVl-15 UFSAR Rev. 14 (June 19961 TYPICAL PENETRATIONS FLASHING SLEEVE PENETRATION SMALL PENETRATION PENETRATION INTERSECTING FORM RIBS NOTE: CUT FLASHING FOR SLEEVE IN FlfLD 2 UNIT PENETRATION ftGURE XVl-16 UF$A1t Rev. 14 (June 1996) REACTOR SHIELD WALL CONSTRUCTION DETAILS Zl'-2 1/2" OIM£TU (lllSlll( or Ill.ATES) z*

• 2-11111* 25'7 1/4" DINUEl (OVTSlll( or Pl.ATES) I Sii" SLEEYE Ill.ATE -tpJ+ ENLARGED PLAN OF TYPICAL COLUMN HORIZONTAL WELD DETAIL S' -6" lllSllJ( Ill SLEEVE SECTION 2-2 SECTION 3-3 ., : ..J ..J c -l w :c (/) 0::: 0 I-(.) <( w 0::: ...J UJ V) V) LIJ > a:: ... c .... ... ' u 421 < ... . " LIJ **.* ... a:: I SECTION 1-1 FIGURE XVl-17 . . *.* . . . : . . ' . I I* .J _J c .J w I (/) n:: (.) n:: ' .... .... ' iii '<t I " ' ,,,. j UFSAR REV. 16 (NOVEMBER 1999)

VENT PIPE AND SUPPRESSION CHAMBER DRYWELL SHELL A>.** .* *:'* 0 I ' .... , * '*! * . *.* . *.: . *,. :* . .. , * .. .. . . :'* ...... ** .. . 6" DIAMETER PIPE SUPPORT MEMBERS : .* ,.* . .... . . . _; .. ... . .. ' ... * .. ii .. : . .. " ** ...... ,. .. . . . * . ....... . .... "" .. =. .. -_-.-. ""!':-. -.-.-.-** -.-,-. -*** -.-. ** -.,-. -. -. -.-.-.-.......... -. -.-. ........ ,.: ** . . .*,.,. .... . . . .. . .: . * . *. ... . ....... . ... ' .. * ... : > *:*** : . .,:: . : .. *. :*: ... *: *:. . .. . . .* *. .. : * .. * .. : . RGURE XVl-18 UfSAR Rev. 14 (June 1996) PRIMARY CONTAINMENT SUPPORT AND ANCHORAGE ..*... ELEVATION PLAN DETAIL A ELEVATION PLAN DETAIL C . .... * ... ELEVATION DETAIL B AGURE XVl-19 UFSAR Rev. 14 (June 1996)

DRYWELL SLIDING -ACCELERATION, SHEAR, AND MOMENT ORYWELL SHELL EL. 225'*6" EL. 214'..S" ----* . . EL 225'-S-' -------*-------.137 1291 . . . ' .. *.* ... * .. EL. 198'.0" . * ....... : . : .... *' ,* ** *** I ----; t,' 1 'C I \ )_ ' > ....__ 1476 EL. 214'*6" '---3981 1755 5891 3665 EL 198'-0" ------L----' .110 208,053 ACCELERATION -g SHEAR -KIP MOMENT -KIP -FT. FIGURE XV1-21 UFSAR Rev. 14 I.June 1991) SHEAR RESISTANCE CAPABILITY -INSIDE DRYWELL --------H DRYWELL SHELL *" :':. :.-.... _ ...... * .. -::* See Figure XVl-24 .. SUPPORT ::.t*.*. SKIRT . EL.214'-6 :: *.: :* EL.212'-0 '* *: .: ,._ . SHEAR CAPABILITY Kips 1 in. x 6 in. plate Concrete 2520 446 75 Nelson studs, 3/4 in. Tota I shear capab i I i*ty shear diameter 931 = 3897 Notes: a. Allowable values based on normal code stresses. Load 1 2 3 b. If a conservative friction factor of 0.3 is assumed, an additional shear capability of 2580 kips for loads 1 and 3 and 5610 kips for load 2 result. c. Loads taken by the drywell skirt. SHEAR LOADS CH) INSIDE DRYWELL AT EL.214'-6 (KIPS) Shear CH> Operating & Earthquake 1476 Load 1 & Flooding to El.333'-0 1476 Load 1 & Jet Thrust 1966 Shear Capability 3897 AGUIE XVl-22 UflAR Rev. 14 t.JwM 1111) SHEAR RESISTANCE CAPABILITY -OUTSIDE DRYWELL : * *4. :*. -: **. :* .! .. , .. qo: .. *t -: .. * .. '/ .. 'EL. 212'-0 ----*H Notes: a. Based on normal code stresses. b. If a conservative friction factor of 0.3 is assumed, an additional capability of 2970 kips for loads 1 and 3 and 5950 kips for load 2 result. SHEAR LOADS CH) OUTSIDE DRYWELL AT EL.2141-6 CKIPS> DRYWELL SHELL . :.; :.:*; ,.:-:. See

• XVl-24 *t. Load Shear CH) Operating & Earthquake 1755 2 Load & Flooding to El.3331-0 3981 3 Load 1 & Jet Thrust Shear Capabl llty 2245 8700 FIGURE XVl-23 UFIAR Rev. 14 (June 1996)

( DRYWELL -SUPPORT SKIRT JUNCTION STRESSES TENSION COMPRESSION B SHEAR SUPPORT SKI RT .** .. : .. 1 .09" DRYWELL SHELL TENSION : 1 /2" .. ;, *. 201-1" R. 53 PSl=ERECTION SEE NOTE a. ; ",4 _;..

• SEE NOTE a. Loads Cps i) 1 2 3 Operating & Load a. & Load a. & Earthquake Flooding Jet Thrust Shear A 2000 2000 2650 Compression 8 4279 8730 5354 Tension B 3186 18,850 4261 Shear C 2620 5900 3350 NOTES: a. Al I vertical loads taken directly to concrete after concrete placement b All horizontal loads taken by support skirt to concrete after concrete placement c. Stresses resulting from Loads 1 or 3 d Stresses resulting from Load 2 4 5 Erection Allowable 5450 1367 5400 c. 16,000 d 19,250 13,000 RGURE XVl-24 UFIAR Rev. 14 (June 1191) 78 88 89 POINT LOCATION FOR CONTAINMENT SPRAY SYSTEM PIPING HEAT EXCHANGER TO DRYWELL HEAT EXCHANGER 42 '52 53 4139 PSI (THERMAL)< 63 116 *115 / 17012 PSI (STATIC E. Q.) 114 17913 PSI (DYNAMIC E. Q., 41%) 113 DRYWELL PENETRATION X-150 FtQURE XVt-21 UFSAR Rev. 14 (June 1996)

COMPARISON OF STATIC _AND DYNAMIC STRESSES (PSI) SEISMIC CONDITIONS CONTAINMENT SPRAY SYSTEM HEAT EXCHANGER TO DRYWELL 10' , "' ... ---\ _...y v 'i_....... \.::: .... -...:::.:: .... , l .J .,.. ' ,___ 103 0 10 20 30 50 I: l .-1

• I* ;'l ' I \ */ -.1 A\ *" .' \\ 17 lJ \ --. --.'/ DYNAMIC (4Jr.i// I Vf ' ' I 11 ---I I .. 60 70 80 90 100 110 120 130 150 POINT NUMBER FIGUN: XVl-28 UfSAR ........ 14 1"'-1996)

CONDUCT-ION IN A DROPLET ( 80 I 70 60 50 \ .. \ \ \ \ 30 I\. ""' "' ' 20 r----.. """---10 0 0.2 .OA 0.6 0.8 1.0 / ftQURE xvt-27 UFIAR "8v. 14 Uune 1-) 70 60 50 40 ........_ ' ..... 30 ... ... '-..... v .. /su.PPRESSION CHAMBER , 1 10 I I I I I 1111 I I I I I INITIAL PRESSURE = 0 PSIG IN ITIA.L WATER TEMPERATURE = 90F DOWNCOMESUBMERGANCE =* 4*FT. ,.,..--' ---\ '"'" "'--... CASE 1 "' ""I ' CASE 2 .... "" -"" "'" ,...., ............ """"' ... _ CASE 3 CASE 4 102 103 104 . TIME AFTER ACCIDENT, SEC ........ 105 5 V> V> 0 .,, n 0 0 z -I > n n -c m z* -I n 0 z -I > z == m z -I .,, :;:a m V> V> c: :::a m 70 60 50 ,_ ...... .._DRYWELl:. .... """" --40 -......... -... --"' "'.,. / 30 CHAMBER I f -I o -1 10 I I I I 11111 I LIMITING INITIAL CONDITIONS INITIAL PRESSURE \, j I I I I = 2 PSIG INITIAL WATER TEMPERATURE = l45F DOWNCOMERSUBMERGANCE = 5 FT. -",, -\ .,... -"" \ ' '-....... ..... .. ..... m2 ro3 10' TIME AFTER ACCIDENT, SEC ----b U> U> 0 'T1 n 0 0 > z -t > n n c m z -t n 0 z -t > -z ' s: m z -t "'O ::0 m U> U> ' c: ::0 m , / NOZZLE SPRAY TEST -PRESSURE DROP OF 80 PSIG AOURI XVl-30 UFSAR Rev. 14 (June 1991) NOZZLE SPRAY TEST -PRESSURE DROP OF 80 PSIG FIGURE XVl-31 UFIAR Rttv. 14 (.Jwl.e 1191) NOZZLE SPRAY TEST -PRESSURE DROP OF 30 PSIG FIGURE XV1-32 UflM ""'* 14 (June 1991) NOZZLE SPRAY TEST -PRESSURE DROP OF 30 PSIG FIGURE XVl-33 Uf&AR ... ¥. 14 (June 1991) ( "' EL.384:0" * .9 " EL.340'*0* * .9 d) !: E:L.318'-o"

• 0 o I -EL.2qe'-0" N -.9 EL.281:o** g
• EL.2":o* I 0 .. EL.237:o* -0 *I * .9 q-* 0 *I N N "o .. 2 :<;> t:: .. 0 .. 2
• 0 . . N b
• I co SEISMIC ANALYSIS -REACTOR BUILDING OPERATING FLOOR 50'*0" 120 118 DEFLECTION (Mils) I J % Ground Motion 584 555 MATHEMATICAL MODEL FIGURE XV1-34 UFIAR Rev. 14 (June 1996)

( ." DYNAMIC ANALYSIS -DRYWELL EL.303'*3** EL. 222 :.<;" MATHEMATICAL MODEL Flooded to El.333'-o* 0 120 DEFLECTION (Mils) FIGURE XVl-31 UFIAR lllev. 14 (June 1191) uJ I-f--0:: ll.l 0 ct a..\,) a.. z =>o V>t) REACTOR SUPPORT STRUCTURE -SEISMIC :r V) MATHE.MATICAL MODE.L R/y BA5E. EL.259!..S* Deflection Result Jet Thrust Seismic Reactor Bldg. Total DE.FLE.CTI ON CURVE. R/V Base Shield Wall R/V Top of El. 259'-5" El. 303'-3" El. 3031 -311 _JY::!_ 13 .13 100 120 0 25 114 159 0 44 59 62 13 82 273 341 DE.fLE.CTION CHART (In Mi Is) FIGURE XVl-:M Uf&All Rev. 14 I.June 19HI ( ' EL. 278'-0" EL.24G'..G" EL.227:"" ..._: . (.9 Q 0 *o r...: C() tri -b .. Ln -t:i C'i C" --U') in . -SEISMIC ANALYSIS -WASTE BUILDING 8 100' E-W DE FLECTJON (Mi Is) 11 % Ground Motion MATHEMATICAL MODEL f1GURE XYl-37 UF&AA Rev. 14 Wune 1191) SEISMIC ANALYSIS -SCREEN HOUSE EL.26f-O" 7 JI so.o* .so.o* MATHEMATICAL MODEL E-W DEFLECTION I I 3 Ground Motion FIGURE XVl-38 UflAR Rev. 14 (June 1996) --( ' SEISMIC ANALYSIS -TURBINE BUILDING (NORTH OF ROW C) \P

• g *' -er 'c0
• co . . .. .... .... EL.3'q:o* *o
• Cf) *;.... .. "' -er . er .. -N 'o * *' 0 tr .. Q cO -C:o .. -EL.31q:.4 er t't1 . *g .... .. EL.505!.G * * .'P \Q en .. 'O Ill *o :9 EL. *:s er Q *o tr -* .. E.L.277*0' 'o in .. co E.L. b
• 0 -* -* er !1 * .9 MATHEMATICAL MODEL (North of Row *c ") N-S E-W DEFLECTION (Mi 11 % Ground Motion FIGURE XVl-39 UflAR Rev. 14 (June 1998)

SEISMIC ANALYSIS -TURBINE BUILDING (SOUTH OF ROW C) E:L.333:0" b b .. *' -er d) -EL.320!0° a 0 -* 0 N EL.500:0" -----+--t-t 4 'I 0 *ca 0 *' W) an N EL.277:0" *3 0 ' 0 -. *I Ln \D -EL2bl-0. 2 : 0 * -' 0 er .. co EL.245*o* MATHEMATICAL MODEL (South of Row'"c' DEFLECTION I I% Ground Motion FIGURE XVl-40 UfSAR .._.v. 14 (June 1991)

---

1 MATHEMATICAL MODEL EL.350'-ll" EL.270'-1 I" G2 4 DEFLECTION (Mils) 11 % Ground Motion 7. 5 % Dampinq n 0 z n ,, m -i m < m z "I -i 0 z Vl -i > n "'

REACTOR BUILDING MATHEMATICAL MODEL (NORTH -SOUTH) EL.384'0" -------<12 MATHEMATICAL MODEL 5% Damping 2.5% Damping in Rock Node Point 6 5 4 3 2 7 12 Walls Walls Floor H20 rnk Floor Above Below L.L. + Wt.k Wt.k Wt.k Wt.k Equip. 4,379 5,017 1,000 1,975 12 ,371 4,431 5,017 5,367 735 2,035 17,585 5,105 5,367 4,796 920 626 16,814 4,916 4,796 6,430 769 16,911 4,989 6,440 8,897 668 20,994 8,996 37,923 906 47,825 37,102 3,860 *13,095 55,075 735 1,175 1,910 *Includes Reactor Support, Drywell, and Pressure Suppression Chamber Area I Shear Elem From To (Ft.2) (Ft.4) A.(Ft.2) 1,880 8,530,000 1,575 2 3 1,420 4,900,000 1,182 3 4 1,260 2,350,000 1,050 4 4 5 1,620 3,640,000 1,350 5 6 1, 160 3,408,000 970 6 7 7,900 23,200,000 6,584 7 8 7 80,000 230,000,000 66,000 8 7 10 80,000 230,000,000 66,000 9 9 8 148 10 ll 10 148 11 6 12 AGURE XVl-42 UHAR lllev. 14 I.June 19961 ..J w <I) \I') ILJ > 0::: 0 u <( lU a:: uJ 1-1-lJJ a..<.) a. z :::>o VlU REACTOR SUPPORT STRUCTURE -SEISMIC :::c:: \/) EL. 225'-G" MSE. CONCRETE MATHE.MAT/CAL MODE.L DEFLECTION (Mi Is) 159 -2, 850 G,237 .___ ______ ___. .... -3C:., 701 MOMENT (Kip ft.) AGUftE XVt-43 UFSAR Rev. 14 (June 19HI _J LU <I) U) lL.I > a'. 0 1-<J <C UJ 0:: uJ o!uJ 0 ct a..u a.. z :::>o .,,u REACTOR SUPPORT STRUCTURE -REACTOR BUILDING 72 MILS ::i: V) E'.L. 225'-<;,' MSE CONCRETE MATHEMATICAL MODE.L DEFLECTION (Mils) 0 -<D,990 L-.---.L -17,750 MOMENT (Kip ft) FIGURE XVl-44 UFSAR Rev. 14 (June 1996) -' w Ill If) LU > a:: 0 ..... <.) < UJ O! u.J a!w 0 ct a.v a.z ::::>o "' CJ REACTOR SUPPORT STRUCTURE REACTOR BUILDING & SEISMIC I V') EL. 225'-<;,' 5ASE. CONCRETE MATHE.MAT/CAL MODE.L DEFLE.CTION (Mi 1&) MOME.NT (Kip ft.) RGURE XVl-45 UFIAR Flev. 14 I.June 19961 PLAN OF BUILDING I 0 I . O'I co l'I . ..:i w EL. 393*-6* EL.397'-l 1/8* ------1 REACTOR BUILDING EL. 393'-6* EL. 333*-o* EL. 364'-4 7/8* TURBINE BUILDING I ID ' ,.., Ill I . 0\ co M . ..:i w I 0 I 0\ co M . ..:i w SCREEN AND PUMPllOUSE WASTE DISPOSAL BUILDING EL. 289'-0* DETAIL *D* DIESEL 1---------EL. 368 '-0 1/4* GENERATOR BUILDING I EL. 364'-4 7/8* 0 I 0\ O'I M EL. 333*-o* . ..:i w EL. 2161-6* EL. 289'-4. ADMINISTRATION BUILDING* SHOPS AND STORES EL. 276'-4* N

REFERENCE:

WALL 11 -P. XVI-170, FIG. XVI-47 P. XVI-171, FIG. XVI-48 P. XVI-172, FIG. XVI-49 P. XVI-173, FIG. XVI-50 WALL 12 -P. XVI-176, FIG. XVI-53 WALL 13 -P. XVI-177, FIG. XVI-54 P. XVI-178, FIG. XVI-55 WALL 13A-P. XVI-178, FIG. XVI-55 WALL 14 -P. XVI-179, FIG. XVI-56 P. XVI-180, FIG. XVI-57 P. XVI-181, FIG. XVI-58 WALL 15 -P. XVI-182, FIG. XVI-59 WALL 16 -P. XVI-183, FIG. XVI-60 WALL 17 -P. XVI-184, FIG. XVI-61 DETAIL *D* -P. XVI-174, FIG. XVI-51 DETAIL *g* -P. XVI-175, FIG. XVI-52 RGURE XVl-41 UFIAR Rav. 14 (JuM 1998) WALL SECTION 1 I ' 2" EXP. JT. EL. 393'-6" 14WF30 SEE DETAIL "A", PAGE XVI-171 Ll.L = 1.33" EL. 383'-8" NOTE: SEE WALL SECTION EL. 383'-ll 3/4" TOP OF RAIL EL. 372'-0" 175:/IRAIL 24!90 EL. EL. 340'-0" EL. 318'-0" EL. 298'-0" EL. 2s1*-o* 14WF34 14WF30 ' 12c20.1\ EL. 369'-0" SHEET 3 OF 9 1. FOR END CONDITION AT 2. COL. LINE 12, SEE DETAIL "D" FOR END CONDITION AT COL. LINE 4, SEE DETAIL "E" ..,_,.-,..------------i"*J e" SLAB 18WF45 -SEE DETAIL "B", PAGE XVI-172 Ll.L=0.93" SEE DETAIL "B", PAGE XVI-172 Ll.L=0,59" TURBINE ) :. ** '. v I _ BUILDING *. a. 12 20:7\ EL. 333'-6" 18WF45 8 SLAB *SEE DETAIL "C", PAGE XVI-173 Ll.L= 0.17" EL. 319'-4" Ll.L =COMPUTED MAXIMUM HORIZONTAL DEFLECTION EL. 305'-6" e'SLAB I SEE DETAIL "C", PAGE XVI-173 Ll.L= 0.13" EL. 291'-0" J 811SLAB SEE DETAIL "C", PAGE XVI-173 Ll.L=0.13" EL. 277'-0" 18WF45 :::I 8" SLAB DETAIL "C", PAGE XVI-173 Ll.L= o.oe" l" EXPANSION JOINT FILLER "FLEXCEL" AS !!!ADE BY CELOTEX CORP. EL. 261'-0" TYPICAL WALL SECTION BETWEEN TURBINE BLDG. AND REACTOR BLDG. (L.OOKlttG EAST) FtGURE XVl-47 UFSAR Rev. 14 (June 1991) WALL SECTION 1 -DETAIL "A" ... (j) K z : 1 '-4 l'-9 x .. BUTYL EXPANSION JOINT AS MADE BY "..,} 16 GA. ALUMINUM FLASHING 3 PLIES 15' FIN. FELT CARLISLE TIRE & RUBBER CO-, FASTENED 1 SIDE 1 PLY 43# BASE FELT .,.._. mm 211 RIGID INSULATION I NKX

• 18/16 GA. STEEL Q-OECK AS MADE .
• EY H. H. ROBERTSON CO. l. CONT. SEALANT -i TOP /STEEL 14 GA. CLOSURE 14 GA. CLOSURE EL. PLATE PLATE F*IE, 11/211 x 18 GA. STEEL FKX
• GA. STEEL FACE PANEL AS MADE BY H. H. ROBERTSON r:o. '* '* ""'""" t=' 4* 1/21 TIIK. LAYERS, U.L. RATED 14 WF 30 : rniPSUM SOARD (2 HH. FIRE HATED) 14 WF 30 -.,8 7/811 NOM. REACTOR 11-4 1/4 4 s;a" I 1/8 TURBINE BLDG. I BLDG. 3'*3 DETAIL "A" FIGURE XVl-48 UFIAR Rev. 14 (June 19H)

( K WALL SECTION 1 -DETAIL"B" 11*4 l/4 JOINT 4 5/811 11-6 1/8 BK./ 12 ( CURB -----------.

• 1/411 THK.
• t. (CONT.) 14 GA. CLOSURE PLATE F *IE, I 1/211 x 18 GA. ST!E L FACE PANEL AS MADE BY H. H. ROBERTSON CO. 4* IJ' THK. LAYERS. 11.L. RATED GYPSUM BOARO '(2 HR. HATEDJ 8 7/8" NOM.
• I DETAIL "B" J !L. VARIES -----. r-*-= I
• a* SLAB I . .
• 18 WF 45 FIGURE XVl-49 UflAR Rev. 14 (June 1996)

REACTOR BLDG. REACTOR BLDG. WALL SECTION 1 -DETAIL 11C11 y 3 ,. -6 , .. *, ... . . FOUNDATION WALL . . 3' -3 . . . . . 1'-6 '1!' CLEAR BK./12[ CURB 12C 20.7 DETAIL *c* 18 WF 45 THIS CONDITION OCCURS AS NOTED BELOW: FLR. EL. 333'*8 BETWEEN COL'S. JlO THRU Jl2 FLR. EL. 305'-6 !!ETWEEN COL'S. J4 THRU Jl2 INCL. FLR. EL. 291'*0 BETWEEN COL'S. J4 THRU J12 INCL. FLR. EL. 277'-0 BETWEEN COL'S. J9 THRU Jl2 INCL. NOTE: COL. BAYS 20'-0 C/C K 3'-3 1'..g J: .. * ... .. '. * .. ( . * .. . .. : . *.* ... :, . : : ,. : : . .-.. .. : 1 .. ... *r* '* '; .. * .. .,.. 1" 1'*5 , .. f . ** J EL. 277'-0 "4. .* ... . ... .. : . * ... .. .. *: . .(_ *:. I" EXPANSION JOINT FILLER ' .. . .. ... . ... : .. *." . *' .... ' . . . ,.* ... :. . *.. ' .. " . .. .... ...... WALL SECTION BETWEEN COL.'S 0 THRU 0 (LOOKING EAST) FLOOR EL. AS NOTEO TURBINE BLDG. TURBINE BLDG. FIGURE XVl-50 UFSAR Rev. 14 (June 1191} K 11-41/4 f ..a I 5/811 3'*3 .... z g PLAN DETAIL "D" !:::.. L = 1.33 16 GA. METAL COATED STEEL CLOSURE PLATES, SET IN MASTIC AND BOL; TED F*IE, 11/211 l 18 GA. STEEL ,; FACE PANEL AS MADE BY a PC 16 WF 36 H. H. ROBERTSON CO. I 16 GA. METAL COATED STEEL CLOSURE PLATES, SET IN MASTIC AND BOLTED 11-i 1/8 N--=::t--16 GA. METAL COATED STEEL CLOSURE PLATES, SET IN MASTIC AND BOLTED 14WF COL /:::,. L = COMPUTED MllXIMUll HORIZONTAL DEFLECTIOH == > r r (/) m n -t 0 z .... 0 m -t > r = c = _'? f -I -REACTOR+TURBINE BLDG. BLDG. J 'l1 WF84 L-4 x 6 x 3/811 f*2, 1 1/211 x 11 GA. ALUM. FACE PANEL AS MADE BY H. H. ROBERTSON CO. _, _, 1fo :I! _,o :::> r l; Offi ..Jm lalO I I I 3*.3 PLAN DETAIL "E" = 1.33" _, ... t!!l ol-(§.,... ... ... _, 01-0 m I :ii t! ... c i= * !!I I' *I!, 1 li211 x 18 GA. STEE.L FACE PANEL AS ti.ADE BY H. H. ROBERTSON CO. 16 GA. METAL COATED STEEL CLOSURE PLATES, SET IN MASTIC AND BOLTED -J N-=t---14 WF COL. 4 FKX*Q PANELS INDICATED DOTTED ARE HORIZONTAL BETWEEN EL. 340'--0 & EL. 383'-I FKX*Q-PANELS AS SHOWN, VERTICAL BETWEEN EL. 383'-8 TO TOP OF PARAPET EL. 398'-6 6 L =COMPUTED MAXIMUM HORIZONTAL DEFLECTION \ ) :e > r r en m n -f -0 z .... c m -f > r rii = EL. 311'.0 REACTOR BLDG. EL. 291'.0 REACTOR BLDG. EL. 281'.0 . : l!:L. 261'.0 EL. 314'*0 DETAIL 1 AL= .15" TURBINE AUXILIARY EXTENSION BLDG. AL= .00 1" EXPANSION JOINT l'ILLl!:ft '"l'LEXCEL" AS MADE BY CELOTEX CORP. WALL SECTION 2 REACTOR BLDG. . *4 . .., . . . . 4 . : . : .*

• 4 . 4 *-.1 4"
• 4 . FOUNDATION WALL 1'*2 1'-6 s .. N : ::.. DETAIL 1 l'-4 F2 -1 117'
• 16 GA. ALUMINUM l'ACE PANl!:L AS MADE BY H.H. ROBEllTSON CO. 3 PLIES 15' 1'1N. l'l!:L T & 1 PLY 43# BASE FELT 7' RIGID ROOF 14 Wf 30 TURBINE AUXILIARY EXTENSION BLDG. EL. 289'*8 : .,. : ": *.: >* _:. ';** ...... " .* "MONO LASTO -Ml!:lllC SEALANTASMADEBY .' * *.-4: *.: .* : * *.
• Ml'G. CO. * * :* ." * * * * .L.;;._ < * * "l'LEXCEL" JOINT FILLER *. *'. 4 : *
• AS MADE BY CELOTEX CORP. :°..". ., REACTOR ......E.,-SET IN BLDG. ---c= MAmc r '°" '°
• 1" CLl!:AR FOR EXPANSION 1'-6 1'-4 DETAIL 2 Tltll condltiall OCCllll .._ Col'l-NC-12A Tin PB-12A 16 Wf 36 TURBINE AUXILIARY EXTENSION BLDG. /::. L = COMPUTED MAXIMUM HORIZONTAL DEFLECTION FIGURE XVl-53 UFIAR Rev. 14 CJune 1991)

WALL SECTION 3 1'-9" !'-1Y2" REACTOR BUILDING "' ., -. . ,* EL. 281*-o* ,;, )' "":' V ... ,o,.,, \T _., *. *. q\' ' ,.,, .. . .. ** I>. *d ... .... *

• d .,.*:* . .. . ,. ,. ,.. . . ,. .*, 0. : ** . , ... (' **, .. ' .. 0 .,,. ... ' 0 TOP or STEEL EL. 289*-o* 14WF30 SEE DETAIL 1, PAGE XVI-178 .6.L= .09* TURBINE AUXILIARY EXTENSION BUILDING 1* EXPANSION JOINT FILLER *FLEXCEL* AS MADE BY CELOTEX CORP. AL= COMPUTED MAXIMUM HORIZONTAL DEFLECTION TYPICAL WALL SECTION COL'S. QA-llA THRU QA-12A (LOOKING WEST) FIGURE XVl-M Rev. 14 (June 19")

WALL SECTION 3A -DETAILS Q 41*101/2 11*9 31*1 l/2 ________ ....... ___________ _ ...... ---i-t; 8: d I REACTOR BLDG. ... : .. *4 . . . * (,) fl) co ., . . . .. :E ci; c .J IL. "=co ::i : JC u F2, 1 1/211 x 16 GA. ALUMINUM FACE PANEL AS MADE BY H. H. ROBERTSON CD. 111 CLR. *4 REACTOR BLDG. .... .. *

• 3PLll!:S15# FIN. FELT & 1 PLY 43# BASE FELT ....... 1112 TURBINE AUXILIARY EXTENSION BLOG. 14 WF 30 DETAIL l . :..* . . . FOUNDATION WALL .. . . * * .. I F*2, 1 1/211 x 16 GA. ALUM. FACI!: PANEL AS MADE BY H. H. ROBERTSON CO. 3 PLIES 15 # !'IN. l'EL T ' I PLY 431,llASE !'[LT . , ..
• s TURBINE AUXILIARY EXTENSION BLDG
• 4* .. " ... .. s ::: 14 WF 30 =.09" DETAIL WALL 3A WALL SECTION AT ROOF BETWEEN COL'S. PB-12A THRU QA 12A LOOKING NORTH L = COMPUTED MAXllAUM HORIZONTAL DEFLECTION FtGURE XV1-&I UFSAR Rev. 14 (June 19M)

WALL SECTION 4 PB 1* EXPANSION JOINT IN ROOF DECK WASTE DISPOSAL BUILDING 13/a" EL. 277*-o* 8 Yi II SLAB L . . b * * * * . . . * ,. IP -THIS SLAB CONDITION P OCCURS BETWEEN COL'S PC-15 TO PC-16AA ; '" ... v 9" 9" *v c I

• 0 " *' .*
• 21-911 ..... ' PD TOP OF STEEL ELlt. 289'-0" SEE DETAIL 1, PAGE XVI-180 AL=.60" TURBINE AUXILIARY EXTENSION & SCREEN & PUMP HOUSE EL. 273'-6 1/2" SEE DETAIL 2, PAGE XVI-181 BETWEEN COL. LINES PC-15 TO PC-16 .O.L = .30111 "MONO LASTO-MERIC" JOINT SEALANT AS MADE BY TREMCO MFG. CORP
• 11-611 I" ..... '! ... *o .. *r; " l" EXPANSION JOINT FILLER "FLEXCEL" AS MADE BY CELOTEX CORP. EL. 261'-0" TYPICAL WALL SECTION BETWEEN COL'S PC-15 THRU PC-18 .0. L =COMPUTED MAXIMUM HORIZONTAL DEFLECTION EXCEPT AS NOTED (LOOKING WEST> FIGURE XVl-H UFSAR Rav. 14 (.June 1998)

( TOP/STEEL EL. 289'*0 l" CLEAR EXP. JT. 16 WF 40 WASTE DISPOSAL BLDG. 13/S" WALL SECTION 4 -DETAIL 1 7 5/8" 8 WF 24 18 WF 45 TURBINE AUXILIARY EXTENSION & SCREEN & PUMP HOUSE DETAIL 1 FIGURE XVl-67 16 WF 36 UFIAR Rev. 14 (.June 1991) ( , / WASTE DISPOSAL BLDG. WALL SECTION 4 -DETAIL 2 9" * .. * .. . . *

• FOUNDATION WALL * * * ..
• I . * . . *
• I" EXPANSION *JOINT "MONO LASTO -MERIC" JOINT SEALANT AS MADE llY TRDICO lll'G. CO. I" EXPANSION JOINT "FLEXCEL" AS MADE llY CELOTEX CORP
• EL. 213'.
• vr * *
• 8" SLAB SLIP . JOINT* .. F . fKx
• 16/16 GA. STEEL 6 *
• AS MADE BY H. H. ROBERTSON CO. . L*6 x 3 1/2 x 318 * / . . .. ,,,,,, . . -. . . 4 *"
• 1*-0 * * * * * * . . DETAIL 2 ... TURBINE AUXILIARY EXTENSION & SCREEN & PUMP HOUSE FIGURE XVl-58 UFSAR Rev. 14 (June 1996)

( TURBINE AUXILIARY EXTENSION MOHO LASTCHllERIC JOINT SEALANT AS MAOIE BY TR!MCO MFG. CO. WASTE DISPOSAL BLOG. WALL SECTION 5 IEL.2611-0 ........ .. . . .. . . . .. .. ... : **** t ...... TAR PAPER JOINT * . . *** ... *:*.*. TYPICAL WALL SECTION BETWEEN COL'S J-14B THRU PB-14B (LOOKING NORTH) 21WF55

• 5/111 311 11* 5 3/811 DETAIL 1 4*PLY ROOFING *.,* .. . ,; * .... : ". '12" NOM. CONC. BLOCK . ... : 11 5/8" 6 L = COMPUTED MAXIMUM HORIZONTAL DEFLECTION FIGURE XVl-59 UFIAR Rev. 14 (June 19M)

( FKX

• 16/16 GA. STEEL. Q-PANEL. L.INER AS MADE BY H. H. ROBERTSON CO. EL.. 2891-8 TOP/STEEL. EL.. 289'-0 WALL SECTION 6 1 '-0 12 [ 20.7 18 WF, 21 WF & 24 WF (BEAM VARIES) 4 5/811 3" 1 1/211 GLASS FIBER INSULATION CONFORMING TO FEO SPECS. HH-1-521C. TYPE I, CLASS A F*2 AL.UM. FACE SHEETJI l/211DEEP x 16 GA.) AS MADE BY H. H. R BERTSON CO. 211 THK. RIGID BOARD INSULATION L.3 1/2 x 3 1/2 x 3/8 CONT. SHEL.F NOTCHED AT ROOF PURL.INS 1/411 STAINL.ESS STEEL. SCREWS I 9 O.C. NKX
• 18/16 GA. STEEL. Q-OECK ROOF PANEL.$ AS MADE BY H. H. ROBERTSON CO. TYPICAL WALL SECTION AT EL. 289'-o BETWEEN COL'S. PB-12A THRU PB-148 LOOKING WEST FIGURE XVl-IO UFSAR Rav. 14 (June 1996)

.. / WALL SECTION 7 EL. 2871-0 *' ..... . -... *. * ... ... : ...... * *.* 0 *,'*".: :-, * *" .' I* REGENERATION ROOM .... . . . . 4 ". . ... : .. * .* . . '.'" *. : .... EL. 2751-0 81/211 SLAB WASTE DISPOSAL BLDG. t:L. 2871-0 " : * .. 1.(1 .*.... : .... *, .. ... TYPICAL WALL SECTION ON COLUMN LINE J-K BETWEEN COL'S 15 & 17 LOOKING WEST 1 '-0 3/16 GIRT LINE . . ' .... . .. ., ,., . * . "** t . ... *, *.* .' . : .. v * * ... * * ** 011-6 . : * . .. }' . * ..... : ,. I 1/211 GLASS FIBER INSULATION CONFORMING TO FEO. SPEC. HH-1*521C. TYPE 1, CLASS A I 1'*2 FACE SHEET (ALUM.) (1112" DEEP l 16 GA.) AS MADE BY H. H. ROBERTSON CO. 2* THK. RIGID BOARD INSULATION I 4 PLY ROOFING NKX 18/16 GA. Q-DECK ROOF PANELS AS MADE BY H. H. ROBERTSON CO. 1* : I DETAIL l 6, L = COMPUTED MAXIMUM HORIZONTAL DEFLECTION FtGURE XVl-81 Ufl.M Rev. 14 (June 1996)

NMP Unit 1 UFSAR LIST OF EFFECTIVE FIGURES SECTION XVII Figure Revision Number Number Section XVII EF-XVII-1 Rev. 25, October 2017 XVII-1 15 XVII-2 15 XVII-3 15 XVII-4 15 XVII-5 15 XVII-6 15 XVII-7 15 XVII-8 15 XVII-9 15 XVII-10 15 XVII-11 15 XVII-12 15 XVII-13 15 XVII-14 15 NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section XVII XVII-i Rev. 25, October 2017 SECTION XVII ORIGINAL ENVIRONMENTAL STUDIES A. METEOROLOGY 1.0 General 2.0 Synoptic Meteorological Factors 3.0 Micrometeorology 3.1 Wind Patterns 3.1.1 200-Ft Wind Roses 3.1.2 Estimates of Winds at the 350-Ft Level 3.1.3 Comparison Between Tower and Satellite Winds 3.2 Lapse Rate Distributions 3.3 Turbulence Classes 3.4 Dispersion Parameters 3.4.1 Changes in Dispersion Parameters 4.0 Applications to Release Problems 4.1 Concentrations From a Ground-Level Source 4.2 Concentrations From an Elevated Source 4.3 Radial Concentrations 4.3.1 Monthly and Annual Sector Concentrations 4.4 Least Favorable Concentrations Over an Extended Period 4.4.1 Ground-Level Release 4.4.2 Elevated Release 4.5 Mean Annual Sector Deposition 4.6 Dose Rates From a Plume of Gamma Emitters 4.6.1 RADOS Program 4.6.2 Centerline Dose Rates 4.6.3 Sector Dose Rates 4.7 Concentrations From a Major Steam Line Break 5.0 Conclusions B. LIMNOLOGY 1.0 Introduction 2.0 Summary Report of Cruises 3.0 Dilution of Station Effluent in Selected Areas 3.1 Dilution of Effluent at the Lake NMP Unit 1 UFSAR Section Title Section XVII XVII-ii Rev. 25, October 2017 Surface Above the Discharge 3.2 Dilution of Effluent at the Site Boundaries 3.2.1 General 3.2.2 Dilution of Effluent at the Eastern Site Boundary 3.2.3 Dilution of Effluent West of the Station Site 3.3 Dilution of Effluent at the City of Oswego Intake 3.3.1 Tilting of the Isothermal Planes and Subsequent Dilution 3.3.2 Dilution as a Function of Current Velocity 3.3.3 Percent of Time Effluent Will be Carried to the Oswego Area 3.3.4 Mixing With Distance 3.3.5 Oswego River Water as a Buffer to Prevent Effluent From Passing Over the Intake 3.3.6 Summary of Annual Dilution Factors for the City of Oswego Intake 3.4 Dilution of Effluent at the Nine Mile Point Intake 3.5 Summary of Dilution in the Nine Mile Point Area 4.0 Preliminary Study of Lake Biota Off Nine Mile Point 4.1 Biological Studies 4.1.1 Plankton Study 4.1.2 Bottom Study 4.2 Summary of Biological Studies 5.0 Conclusions C. EARTH SCIENCES 1.0 Introduction 2.0 Additional Subsurface Studies 3.0 Construction Experience 3.1 Station Area 3.2 Intake and Discharge Tunnels 4.0 Correlation With Previous Studies 4.1 General 4.2 Geological Conditions 4.3 Hydrological Conditions NMP Unit 1 UFSAR Section Title Section XVII XVII-iii Rev. 25, October 2017 4.4 Seismological Conditions 4.5 Conclusion D. REFERENCES NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section XVII XVII-iii Rev. 25, October 2017 XVII-1 DISPERSION AND ASSOCIATED METEOROLOGICAL PARAMETERS XVII-2 RELATION OF SATELLITE AND NINE MILE POINT WINDS XVII-3 FREQUENCY OF OCCURRENCE OF LAPSE RATES - 1963 AND 1964 XVII-4 RELATION BETWEEN WIND DIRECTION RANGE AND TURBULENCE CLASSES XVII-5 STACK CHARACTERISTICS XVII-6 DISTRIBUTION OF TURBULENCE CLASSES BY SECTORS XVII-7 SECTOR CONCENTRATIONS - 1963 SECTOR A SOURCE XVII-8 SECTOR CONCENTRATIONS - 1963 SECTOR B SOURCE XVII-9 SECTOR CONCENTRATIONS - 1963 SECTOR C SOURCE XVII-10 SECTOR CONCENTRATIONS - 1963 SECTOR D1 SOURCE XVII-11 SECTOR CONCENTRATIONS - 1963 SECTOR D2 SOURCE XVII-12 SECTOR CONCENTRATIONS - 1963 SECTOR E SOURCE XVII-13 SECTOR CONCENTRATIONS - 1963 SECTOR F SOURCE XVII-14 SECTOR CONCENTRATIONS - 1963 SECTOR G SOURCE XVII-15 SECTOR CONCENTRATIONS - 1963 SECTOR A SOURCE HEIGHT GROUND NMP Unit 1 UFSAR Table Number Title Section XVII XVII-iv Rev. 25, October 2017 XVII-16 SECTOR CONCENTRATIONS - 1963 SECTOR B SOURCE HEIGHT GROUND XVII-17 SECTOR CONCENTRATIONS - 1963 SECTOR C SOURCE HEIGHT GROUND XVII-18 SECTOR CONCENTRATIONS - 1963 SECTOR D1 SOURCE HEIGHT GROUND XVII-19 SECTOR CONCENTRATIONS - 1963 SECTOR D2 SOURCE HEIGHT GROUND XVII-20 SECTOR CONCENTRATIONS - 1963 SECTOR E SOURCE HEIGHT GROUND XVII-21 SECTOR CONCENTRATIONS - 1963 SECTOR F SOURCE HEIGHT GROUND XVII-22 SECTOR CONCENTRATIONS - 1963 SECTOR G SOURCE HEIGHT GROUND XVII-23 ESTIMATES OF THE LEAST FAVORABLE 30 DAYS IN 100 YEARS XVII-24 CONCENTRATIONS IN THE LEAST FAVORABLE CALENDAR MONTH - 1963-64 XVII-25 ANNUAL AVERAGE SECTOR DEPOSITION RATES (Vg = 0.5 CM/SEC) XVII-26 ANNUAL AVERAGE SECTOR DEPOSITION RATES (Vg = 2.5 CM/SEC) XVII-27 PRINCIPAL RADIONUCLIDES IN GASEOUS WASTE RELEASE XVII-28 CORRECTION FACTORS TO OBTAIN ADJUSTED CENTERLINE DOSE RATES FOR SECTOR ESTIMATES XVII-29 ANNUAL AVERAGE GAMMA DOSE RATES XVII-30 DILUTION CALCULATION FOR EASTWARD CURRENTS BASED ON WATER AVAILABILITY NMP Unit 1 UFSAR LIST OF FIGURES Figure Number Title Section XVII XVII-v Rev. 25, October 2017 XVII-1 AVERAGE WIND ROSES FOR JANUARY '63-'64 XVII-2 AVERAGE WIND ROSES FOR FEBRUARY '63-'64 XVII-3 AVERAGE WIND ROSES FOR MARCH '63-'64 XVII-4 AVERAGE WIND ROSES FOR APRIL '63-'64 XVII-5 AVERAGE WIND ROSES FOR MAY '63-'64 XVII-6 AVERAGE WIND ROSES FOR JUNE '63-'64 XVII-7 AVERAGE WIND ROSES FOR JULY '63-'64 XVII-8 AVERAGE WIND ROSES FOR AUGUST '63-'64 XVII-9 AVERAGE WIND ROSES FOR SEPTEMBER '63-'64 XVII-10 AVERAGE WIND ROSES FOR OCTOBER '63-'64 XVII-11 AVERAGE WIND ROSES FOR NOVEMBER '63-'64 XVII-12 AVERAGE WIND ROSES FOR DECEMBER '63-'64 XVII-13 AVERAGE WIND ROSES FOR '63-'64 XVII-14 AVERAGE DIURNAL LAPSE RATE JANUARY '63-'64, FEBRUARY '63-'64 XVII-15 AVERAGE DIURNAL LAPSE RATE MARCH '63-'64, APRIL '63-'64 XVII-16 AVERAGE DIURNAL LAPSE RATE MAY '63-'64, JUNE '63-'64 XVII-17 AVERAGE DIURNAL LAPSE RATE JULY '63-'64, AUGUST '63-'64 XVII-18 AVERAGE DIURNAL LAPSE RATE SEPTEMBER '63-'64, OCTOBER '63-'64 XVII-19 AVERAGE DIURNAL LAPSE RATE NOVEMBER '63-'64, DECEMBER '62-'63 NMP Unit 1 UFSAR Figure Number Title Section XVII XVII-vi Rev. 25, October 2017 XVII-20 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JANUARY '63-' XVII-21 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR FEBRUARY '63-'64 XVII-22 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR MARCH '63-'64 XVII-23 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR APRIL '63-'64 XVII-24 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR MAY '63-'64 XVII-25 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JUNE '63-'64 XVII-26 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JULY '63-'64 XVII-27 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR AUGUST '63-'64 XVII-28 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR SEPTEMBER '63-'64 XVII-29 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR OCTOBER '63-'64 XVII-30 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR NOVEMBER '63-'64 XVII-31 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR DECEMBER '63-'64 XVII-32 SECTOR MAP XVII-33 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS I XVII-34 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS II NMP Unit 1 UFSAR Figure Number Title Section XVII XVII-vii Rev. 25, October 2017 XVII-35 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS III XVII-36 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS IV XVII-37 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS II BECOMING CLASS IV AT 2 KM AND CLASS II AT 23 KM XVII-38 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 16 KM XVII-39 CENTERLINE CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 2 KM XVII-40 RADIAL CONCENTRATIONS - TURBULENCE CLASS I XVII-41 RADIAL CONCENTRATIONS - TURBULENCE CLASS II XVII-42 RADIAL CONCENTRATIONS - TURBULENCE CLASS III XVII-43 RADIAL CONCENTRATIONS - TURBULENCE CLASS IV XVII-44 RADIAL CONCENTRATIONS - TURBULENCE CLASS II BECOMING CLASS IV AT 2 KM AND CLASS II AT 23 KM XVII-45 RADIAL CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 16 KM XVII-46 RADIAL CONCENTRATIONS - TURBULENCE CLASS IV BECOMING CLASS II AT 2 KM XVII-47 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS I XVII-48 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS II XVII-49 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS III XVII-50 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS IV XVII-51 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS II BECOMING CLASS IV AT 2 KM AND CLASS II AT 23 KM NMP Unit 1 UFSAR Figure Number Title Section XVII XVII-viii Rev. 25, October 2017 XVII-52 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS IV BECOMING CLASS II AT 16 KM XVII-53 CENTERLINE GAMMA DOSE RATES - TURBULENCE CLASS IV BECOMING CLASS II AT 2 KM XVII-54 ASSUMED CONCENTRATION AND DOSE RATE DISTRIBUTIONS CLOSE TO THE ELEVATED SOURCE XVII-55 GAMMA DOSE RATE AS A FUNCTION OF y AT 1 KM FROM THE SOURCE XVII-56 SOUTHEASTERN LAKE ONTARIO XVII-57 DILUTION OF RISING PLUME XVII-58 ESTIMATED LAKE CURRENTS AT COOLING WATER DISCHARGE XVII-59 TEMPERATURE PROFILES IN AN EASTWARD CURRENT AT THE OSWEGO CITY WATER INTAKE XVII-60 SUBSURFACE SECTION PLOT PLAN XVII-61 LOG OF BORING (BORING CB-1) XVII-62 LOG OF BORING (BORING CB-2) XVII-63 LOG OF BORING (BORING CB-3) XVII-64 LOG OF BORING (BORING CB-4) XVII-65 ATTENUATION CURVES NMP Unit 1 UFSAR Section XVII XVII-1 Rev. 25, October 2017 SECTION XVII ORIGINAL ENVIRONMENTAL STUDIES A. METEOROLOGY 1.0 General Two full years of micrometeorological data was obtained from the tower and satellite equipment described in the Preliminary Hazards Summary Report (PHSR), Volume II, Appendix A. This report summarizes the results of the analyzed data for the years 1963-1964. There is no evidence of any important difference between meteorological conditions during 1964 and those observed in 1963. There are variations, of course, such as slightly different wind direction patterns, but none have been found which have a significant bearing on either the routine dispersion patterns or the estimates of limiting dispersion conditions. For this reason, tower operation was discontinued early in 1965. This analysis is, therefore, confined largely to refinement of the dispersion estimates; the general meteorological factors discussed in the PHSR will not be recapitulated here. The 1964 meteorological studies were conducted in three phases. The first portion was to achieve better definition of dispersion conditions in the immediate vicinity of the site, accomplished primarily by more complete analysis of the local tower data than time permitted in the preparation of the PHSR. The second phase involved consideration of the changes in meteorological conditions that one would expect beyond the first kilometer or two, with particular reference to the transport across land-water boundaries. This portion of the study was accomplished by refinement of the mathematical models used to describe dispersion. In the final phase, the meteorological studies have been interpreted in more sophisticated fashion than in the initial study. The most significant differences between the current results and those of the PHSR have nothing to do with meteorological factors. Rather, they are derived from consideration of an actual stack release of ventilating air at 350 ft instead of a fictitious source fixed at 250 ft aboveground. The additional 100 ft added to the final stack design, together with the NMP Unit 1 UFSAR Section XVII XVII-2 Rev. 25, October 2017 release of the large volume of warm air, makes a very noticeable difference in concentrations and associated effects close to the Station. 2.0 Synoptic Meteorological Factors The description of the general meteorology of the area given on page A-1 of the PHSR remains complete and valid, except for the following. On page A-2 of the PHSR it was suggested that a reasonable estimate of the maximum short-term snowfall was approximately 50 in in a period of 72 hr. The 1965-1966 winter indicates an upward revision of that estimate to approximately 75-90 in in the same period of time. Reports of the very heavy snowfall this past winter reached 100 in, but there was much drifting and probably a bit of local exaggeration in this estimate. In any case, snow depths observed at the site were certainly not as great as those slightly inland. 3.0 Micrometeorology This section is a presentation of the same type of data presented in the PHSR, although it now includes two complete years. Comments are entered only when warranted by some interesting change in either results or procedures. 3.1 Wind Patterns 3.1.1 200-Ft Wind Roses Figures XVII-1 through XVII-13 are 2-yr substitutes for the wind roses presented in the First Supplement to the PHSR, as Figures A-17.1 through A-17.7. 3.1.2 Estimates of Winds at the 350-Ft Level In order to complete calculations of the dispersion of effluent from the design stack, winds at 350 ft aboveground have to be estimated from those measured at the 200-ft tower. In an open area such as this, the assumption of little direction change with height is valid, except in occasional isolated cases. The wind speed, however, does change with height, and the comparison between the 30- and 200-ft levels on the tower has shown that the vertical wind profiles in this area are not significantly different from those observed in most flat areas. NMP Unit 1 UFSAR Section XVII XVII-3 Rev. 25, October 2017 Estimates of the winds at the higher levels have, therefore, been based on a power law approximation of the form: (1) Where: = wind speeds at desired and reference heights = desired and reference heights = exponent, given in Table XVII-1 3.1.3 Comparison Between Tower and Satellite Winds The use of the satellite wind station in Oswego and at a point 1 mi south of the main tower was reported in the PHSR. Since then the satellite station has been relocated to approximately 6 mi south of the site on Whitmore Road. This placement was selected to show whether the tower wind direction was reliable as an indicator of the flow patterns inland. As is clearly evident in Table XVII-2, the relation between the tower and Whitmore Road wind directions is very consistent. The complete study lends confidence that the tower wind is fairly representative of the wind over the entire area, since the data reviewed in the PHSR showed much the same uniformity with different satellite locations. As a matter of interest, the 30-ft instrument on the tower reflected calms much more often than the aerovane at 200 ft used for comparison with the satellite. During simultaneous wind actuation of the 30- and 200-ft instruments, the directions were usually within one compass point of each other. 3.2 Lapse Rate Distributions The monthly lapse rates determined from the 200- and 30-ft temperatures were similar in 1963 and 1964. As with the winds, there are small but not significant variations. Figures XVII-14 through XVII-19 reflect the mean monthly lapse rates for the 2 yr, and Table XVII-3 summarizes the frequency of occurrence of various lapse rate classes. These data provide a good pictorial representation of the general stability structure, and they are especially valuable in drawing attention to the stable dispersion condition over the lake in the daylight hours of the late spring and early summer. 3.3 Turbulence Classes NMP Unit 1 UFSAR Section XVII XVII-4 Rev. 25, October 2017 The plots of the combined 1963 and 1964 monthly distributions of the turbulence classes, Figures XVII-20 through XVII-31, change from 1963. There was a shift between the Class II and Class III turbulence classes, the latter being represented much less frequently in 1964 than in 1963. At first, a classification error was suspected, but a sufficiently large sample of the original records was reprocessed to be quite certain that the shift was real. The indication is that 1964 was somewhat more turbulent and had fewer instances of overcast, neutral conditions. In any case, the change is not very important in terms of dispersion estimates, since the difference between concentrations observed with Class II and Class III conditions is not very large. The mean plots of turbulence class occurrences are believed to be good estimates of the long-term frequency distributions. 3.4 Dispersion Parameters At the time of completion of the PHSR, it was recognized that more attention should be devoted to assuring that the set of dispersion parameters used in the study was typical of the site. This verification has been accomplished by the study of an additifast runs (runs taken at fast chart speed so that frequency distributions of the direction fluctuations can be determined) and the study of the wind direction ranges, which were routinely recorded for each hour of the 2-yr period. The range estimates are particularly interesting since they represent a relatively simple and easy classification system which does not involve the subjective estimation required for the determination of turbulence classes. The percentage frequency of the observed ranges associated with the several turbulence classes thus provides an independent check of the validity and significance of the latter. Table XVII-4 shows that the turbulence classes and range estimates are consistent and meaningful. Most striking is the concentration of the Class IV hours in the 0-9 deg wind direction range group, but that is in accord with the expected poor dispersion conditions of the stable cases. Class I turbulence is associated with wide swings of the wind direction record, and the Class II and III groupings are similar enough in their distributions to justify treating them as a single classification in future studies. Roughly translating the range values into angular standard deviations, one obtains typical values of 12-15 deg for Class I, NMP Unit 1 UFSAR Section XVII XVII-5 Rev. 25, October 2017 5-10 deg for Classes II and III, and a very narrow 1-2 deg for the stable Class IV groups. These studies have shown, therefore, that the Nine Mile Point site has dispersion parameters typical of relatively flat, uncomplicated sites as was suggested in the PHSR. The final set of dispersion parameters selected for the study are reproduced in Table XVII-1. The set derived for the higher elevation, 350 ft, has also been used to represent the ground level release. The layers close to the ground are probably affected by somewhat larger values of in most actual cases, but since the low-level estimates are used only to evaluate any unlikely accident rather than a routine release, a small degree of conservatism will have no practical significance. 3.4.1 Changes in Dispersion Parameters While the parameters presented in Table XVII-1 are considered suitable for describing dispersion under unchanging conditions, it is recognized that there are undoubtedly variations associated with distance, reflecting primarily difference in land and water. It is equally true that a precise definition of these variations is the proper subject for a sophisticated research investigation, but such a study is unnecessary for engineering estimates as long as the mathematical modeling of the problem is conservative. A slight operating penalty may be introduced by deliberately using these conservative estimates in lieu of detailed knowledge. In the PHSR the problem of estimating the possible effects of changes in dispersion parameters with distance was treated by assuming the worst possible conditions as being present over the entire trajectory. This is unquestionably unrealistic. In this study more logical changes are used as the basis for estimating monthly and annual concentrations and doses in the various sectors. The technique followed is that developed by Smith and Singer(1) after lengthy review of the problem with other investigators. This system is more realistic than the simple dispersion estimate in that it takes account of the increase of wind with height and, therefore, produces more meaningful ground-level concentrations. This effect is pronounced only with sources at very low level, where one would expect a rapid change in the wind speed as the clouds expand vertically. On the other hand, this system is more conservative than the simple dispersion estimate by providing a sensible upper limit to dispersion, NMP Unit 1 UFSAR Section XVII XVII-6 Rev. 25, October 2017 rather than assuming that a cloud or plume continues to expand vertically throughout its entire history. This effect is important with either ground-level or elevated sources. Permitting conditions to change with time or distance is believed to be an important feature for Nine Mile Point analysis. This analytical technique allows for such changes. The form of the dispersion equation differs slightly from the Sutton form, and is used primarily for convenience in handling the changes in parameters with distance and time. (2) Where: = concentration (units/m3) R = release rate (units/sec) = crosswind and vertical plume standard deviations (m) = means wind speed (m/sec) at h h = effective stack height (m) x, y, z = downwind, crosswind and vertical distances (m) from source Since the and values are represented as simple power functions of distance for each turbulence class (Table XVII-1), they give precisely the same concentration distributions as the Sutton equation used with the parameters presented in the same table. Variation of the wind with height is accomplished by permitting the value of in equation (2) to increase according to a power function of and a weighting factor that takes account of the vertical distribution of the cloud. This procedure makes the effective transport speed of the cloud increase with distance as long as the vertical extent of the cloud grows. This growth, however, ceases when upward dispersion becomes limited by the presence of an inversion lid aloft in the atmosphere, a phenomenon which is often present at rather low levels. NMP Unit 1 UFSAR Section XVII XVII-7 Rev. 25, October 2017 The selection of values for the inversion lids requires some explanation. The concept of a maximum mixing depth for dispersion is not new, but estimates of the typical thicknesses of such mixed layers are relatively recent. Smith and Singer(1) arrived at approximations of the typical lid heights on the basis of work by several other investigators, since measurements at this site are not available. These general estimates are given in Table XVII-1. In any case the selection has no effect on the concentrations and dose rates within the first 5 km. The ability of the system to treat changes in conditions with time or distance would make it feasible to introduce every possible variation into the estimates of sector concentrations, but it would also entail almost endless subdivision of the data by sector turbulence class and season. (Sector location is shown on Figure XVII-32.) For this reason, turbulence conditions are treated as fixed, except where there is clear indication of a changing condition that might adversely affect the ground-level concentrations. Specifically, these are as follows: Sector C One would certainly anticipate that an unstable flow passing out over the lake in the spring and early summer would quickly become stable and remain so until it again reached land approximately 23 km distant. It is, therefore, assumed that all Class II (typical daytime) turbulence conditions become Class IV as they pass over the lake and return to Class II at the far shore. It is assumed that the effluent initially disperses under Class II turbulence with the horizontal and vertical standard deviations of the plume reaching 245 and 223 m, respectively, at 2 km. From 2 to 23 km the dispersion proceeds according to Class IV conditions with no further vertical growth and a small increase in lateral dimensions. Beyond 23 km, Class II turbulence again dominates, but in order to simulate the fumigation condition described in Hewson in studies of the Fermi Plant(2), the vertical dimension is held constant from 23 km onward. This is a very pessimistic assumption about the lid and certainly results in unrealistically high estimates of concentrations at distances greater than 25 km. Sector D1 NMP Unit 1 UFSAR Section XVII XVII-8 Rev. 25, October 2017 This sector is affected by the same phenomenon, although in this instance the air is already stable, having passed over a considerable stretch of lake before reaching Nine Mile Point. When stable Class IV turbulence is observed during the day, as it often is during the spring, the assumption is made that conditions change to Class II at the shore some 16 km distant. The treatment of lid variations and the fumigation is substantially identical to that of Sector C. Sectors D2 through G Daytime onshore flow into these sectors during the spring may often be stable, and one would expect developing turbulence to cause a fumigation area about 1 to 2 km inland. All daytime Class IV cases are, therefore, assumed to change to Class II as they pass inland. 4.0 Applications to Release Problems Application of the meteorological data to the release of radioactivity from the reactor complex naturally breaks down into consideration of the accident problems on the one hand and the routine on the other. In the PHSR, it was natural to devote roughly equal attention to both, but it is now possible to discuss in depth specific routine emissions. The meteorological estimates of accident situations remain very much as they were, requiring review rather than reevaluation. 4.1 Concentrations From a Ground-Level Source Plots of the typical ground-level concentrations that might be found directly downwind of an accidental release are shown on Figures XVII-33 through XVII-36. Comparison with similar plots shown in the PHSR (Figures A-10.l through A-10.4) shows that the new treatment gives concentrations decreasing more rapidly close to the source as a result of the wind change with height, but the slopes become more gradual as the effect of the lid becomes important. The plots in the new study are different in detail, but essentially the same in terms of a specific accident. The special problems raised by changes in dispersion conditions, as discussed in Section 3.4.1, are reflected in a second series of plots (Figures XVII-37 through XVII-38), each of which applies to certain sectors only. These are most easily discussed in reference to the sectors themselves. Sectors A and B NMP Unit 1 UFSAR Section XVII XVII-9 Rev. 25, October 2017 Dispersion parameters for these sectors were assumed to be constant with distance and time, and representative of flow from the land. Since most of the flow is over water and distances to the nearest land are great, no additional studies were deemed necessary. Sector C Sector C presents special problems associated with the wind trajectory. Typical unstable dispersion conditions found during the daytime hours would be expected to become stable over the water (especially in the spring), and return to instability when the plume reached the shore 23 km distant; Figure XVII-37 represents this condition. The rather complicated centerline concentration pattern gives values at 25 km, approximately 40 times greater than the standard Class II plot. Sector D1 This sector also poses special problems, again largely in the spring season. We have already found that there are many daytime hours characterized by Class IV turbulence (1.7 percent of all hours in the 2-yr average, and 6 percent of the June hours in 1963). This stable flow has apparently already developed over the water before reaching the site. In the model it is assumed that the stable pattern persists until the flow reaches the shore at approximately 16 km, whereupon one would expect a change to typical daytime conditions over land (Class II). Figure XVII-38 represents this condition. Sectors D2 through G This group of sectors may be expected to show a change similar to that at the shoreline in Sector D1 when a stable flow from the water in spring and early summer becomes turbulent over the land. Various studies of sea breeze phenomena suggest that this changeover occurs within the first few kilometers(3) and we have assumed that the transition occurs always at 2 km from the Station. It should be reemphasized that these changeover conditions are especially pertinent to the spring and early summer, but we have chosen to allow them to apply throughout the year. This makes the estimates slightly conservative, but the frequency of stable conditions during the daylight hours in seasons other than spring and early summer is so small that important errors cannot NMP Unit 1 UFSAR Section XVII XVII-10 Rev. 25, October 2017 be introduced. Figure XVII-39 represents this condition for these sectors. 4.2 Concentrations From an Elevated Source The comparison of PHSR and current estimates of concentrations from a low-level source reveal no very startling differences, but the same cannot be said of the stack emission. The curves in the PHSR for the hypothetical 250-ft source and those applying to the real stack in this report differ by a factor of 2-5 in the area close to the site boundary (1-3 km). This difference is entirely associated with the inclusion of the effective stack height in the calculations. First of all, the stack that will serve the facility is 350-ft high, rather than 250 ft, and a rather substantial volume of air will be released at a temperature considerably above ambient. Table XVII-5 shows the estimates of effective stack height based on the Bosanquet et al (1950) equation that has been used in the study. The plots of the ground-level concentrations directly downwind of the elevated source (Figures XVII-33 through XVII-39) are very different from those shown in the PHSR, particularly where the changing conditions are reflected. In Figure XVII-38, for example, the breakup of the stable plume aloft creates concentrations 18-20 km that are approximately 1/3 of those found at 1 km during normal daytime conditions (Figure XVII-34). 4.3 Radial Concentrations On page A-10 of the PHSR a technique was used to develop ground-level concentrations, assuming a constant dispersion condition and a completely uniform wind distribution (throughout 360 deg) during the period in question. This technique was again applied in this report, modified, however (as was done for the centerline concentration plots of this report), to include the variations of wind speed with height and the changes in dispersion conditions with distance. The radial concentrations resulting from this procedure are included as Figures XVII-40 through XVII-46. 4.3.1 Monthly and Annual Sector Concentrations One of the more important by-products of the meteorological analysis was the establishment of mean annual concentration distributions in various directions from the source. This was done in accordance with the procedure on page A-11 of the PHSR. Results were based on the data of Table XVII-6, which presents NMP Unit 1 UFSAR Section XVII XVII-11 Rev. 25, October 2017 the percentage of time that each sector is affected by a given type of turbulence class. The final values derived represent the best current estimate of the annual concentration distribution from a routine stack release. Elevated Source The 350-ft source height sector concentrations are presented in Tables XVII-7 through XVII-14. These tables may be compared to the A-10 and A-11 series of tables in the PHSR and its First Supplement. As one would expect, if the original series were valid first approximations, the differences between the two sets are systematic and easily described. Most obvious is the marked decrease in ground-level concentrations within the first 3 km from the source in all sectors. Also, at the greater distances there are often important changes in the concentrations obtained with the new calculations. In Sector E, for example, the average concentration at 50 km is only 20 percent of that estimated in the original series. This is directly associated with the augmented stack height and the fact that under inversion conditions a plume will seldom begin to reach ground level until it has passed beyond 50 km. It should be noted that a conservative assumption deemed necessary in the PHSR has been deleted from the current analysis. The terrain inland south of the site is rolling, sloping gradually upward with some portions reaching the 250-ft level within 5 km. It was, therefore, assumed in the PHSR that the inversion plume from the 250-ft "stack" would begin to intersect the ground at this distance; all calculations for Sectors E, F and G were based on this concept. In the present study it is recognized that even with a relatively strong wind (10 m/sec) an inversion plume would be approximately 400 ft aboveground. Since the terrain doesn't reach this elevation until somewhat greater than 30 km in Sector D1, it is not at all likely that ground-level concentrations would ever be found closer than this under such conditions. The influence of the more sophisticated analytical system allowing changes in conditions with distance on the long-term concentration patterns, especially in Sectors C and D1, is small. The breakup of the inversion plume as it passes over the land is evident in a decrease in the slope of the concentration curves with distance in the areas where these changes occur, but it is certainly not much more than noticeable. Similarly, the D2 through G sectors show a slight flattening of slope in the 3-4 NMP Unit 1 UFSAR Section XVII XVII-12 Rev. 25, October 2017 km region, but this also is not greatly different from a simple dispersion pattern. Determination of the limiting conditions on an annual basis is clearly associated with Sectors C and D1. Close to the stack the concentrations downwind in Sector D1 are the highest observed, ranging from 10-7 units/m3 at 1 km for a 1 unit/sec release rate to 10-8 units/m3 at 5 km. At very great distances the concentrations on the downwind shore in Sector C create the highest average concentrations, reaching 2 x 10-9 at 25 km and 1 x 10-9 at 50 km. The same is true of these two sectors when one examines the maximums found in individual calendar months. Here the observed peak values are a factor of 2 to 3 higher than the 2-yr means. Ground-Level Source Tables XVII-15 through XVII-22 are similar summaries of results for a continuous ground-level release. These are naturally more like the original set of distributions (A-8 and A-9 series of tables in the PHSR and its First Supplement) than those from the 350-ft source, since the only differences other than the smoothing effect of 2 yr are the increase of wind with height, the capping inversion lid and the change of conditions with distance. In some cases these differences have produced values less than those given in the First Supplement to the PHSR. These new values reflect the more realistic and thorough treatment of the data as discussed in Section 3.4.1. Sector D1 remains the highest on an overall average basis, although not by any great margin, and Sector E is the lowest of the group. The new data continue to reflect the diversity of wind and dispersion conditions in that the range between these highest and lowest sectors is only slightly greater than a factor of 2. 4.4 Least Favorable Concentrations Over an Extended Period In the PHSR a lengthy section was devoted to the problem of estimating the combinations of conditions that might bring the highest possible concentrations to various sectors surrounding the reactor site. As it should be if the original analysis were valid, there is little cause for readjusting these estimates after only one additional year of meteorological data. This is especially true of the ground-level release, where only some new thoughts on the prevalence of certain conditions might serve as NMP Unit 1 UFSAR Section XVII XVII-13 Rev. 25, October 2017 the basis for readjustment, but the elevated data do require consideration of the actual 350-ft stack even though the statistics are unchanged. 4.4.1 Ground-Level Release Table A-12 of the PHSR gives estimates based on synoptic and statistical considerations of the least favorable 30-day concentrations to be expected in a period of 100 yr. Since this table is still considered perfectly valid it is reproduced here as Table XVII-23, and a new Table XVII-24 summarizing actual experience is included for comparison. The upper portion of Table XVII-24 shows the highest calendar months observed for the ground-level releases and it seems clear there that the experience is in accord with the estimates. Sector D1 is indeed the one with the greatest ground-level concentrations, with values at 1 and 10 km approximately 1/2 the statistical estimate of the worst conditions, and 1/3 to 1/5 that for the synoptic evaluation. The discrepancy between estimates and experience becomes greater with distance, largely because the new dispersion model takes into account the variation of wind with height. This is an important consideration for the ground-level source. 4.4.2 Elevated Release The PHSR estimates indicated that one would expect a single set of absolute maximum values as the estimates for all sectors, primarily because one could visualize very persistent (and, therefore, very similar) strong wind conditions occurring in almost any sector. None of the new estimates of concentrations, of course, approach the 0.2-km value of 40 x 10-8 because of the greater height of the source. At greater distances, however, both Sector C and D1 show more realistic values approximately 1/4 to 1/5 of the PHSR estimate. The new 10-km figure for Sector C is greater than 1/2 the PHSR estimate because of the conservative treatment of the fumigation case. On page A-17 of the PHSR, the least favorable conditions for periods less than 30 days were discussed. There it was pointed out that for an elevated source, the synoptic condition producing relatively invariant wind directions would be associated with well-developed, but nearly stationary, pressure systems. It seems clear that for periods as short as 5 days, either Class II or Class III turbulence could predominate to a marked degree, although in longer periods the distribution between the two would probably become more nearly equal. Thus, NMP Unit 1 UFSAR Section XVII XVII-14 Rev. 25, October 2017 the following distributions of conditions are considered as reasonable upper limits. Time Period Wind Direction Turbulence Class 90% Class II 10% Class III 5 Days Confined to 45° or Sector 10% Class II 90% Class III ------------------------------------------------------- 70% Class II 30% Class III 10 - 15 Days Confined to 45° or Sector 30% Class II 70% Class III The data have been searched for the least favorable 10 days that might be associated with a release from the stack. A period was found (August 8-18, 1964) in which the wind was blowing into Sector D1 113 out of a possible 240 hr. Virtually all of this time (110 hr) was of Class II turbulence. This is somewhat more pessimistic than reported in the PHSR. In terms of concentrations close to the site boundary, this period would have produced by far the highest values ever observed in any sector, reaching 27 x 10-8 at 1 km and 4.8 x 10-8 at 3.0 km. These values compare with maximum estimates of 40 x 10-8 and 12 x 10-8 at 1.0 and 3.0 km, respectively, derived from the table above. 4.5 Mean Annual Sector Deposition Deposition has been treated in a very simple fashion, based entirely on the concentration estimates (without corrections of any kind) and deposition velocities. The product of the concentration and deposition velocity yields the deposition rate (in units per unit area per unit time). Tables XVII-25 and XVII-26 show the mean annual deposition rates that would be found with radionuclides having deposition velocities of 0.5 cm/sec and 2.5 cm/sec. These velocities are believed to NMP Unit 1 UFSAR Section XVII XVII-15 Rev. 25, October 2017 represent fairly typical and upper-limit values, respectively, for most of the radionuclides in question. Refinements such as discriminating among specific radionuclides and correcting for the removal of deposited materials as the plume moves downwind would not affect the data significantly in comparison with the basic uncertainties. Corrections and contributions for the washout of the cloud by rain and snow would also have little effect on the data. Englemann et al(4), in their recent studies, have shown that the washout process for very small particles (which we may safely assume to be typical of the Nine Mile Point particulate emissions) is far less important than earlier speculation suggested. 4.6 Dose Rates From a Plume of Gamma Emitters Many of the radionuclides routinely released in small quantities through the tall ventilating stack will create radiation at ground level even though the plume itself may be aloft. This problem has often been calculated in early safety analyses by very rough approximations, since accurate representation was an impossible task. Recently, computer programs have been developed that reduce the effort to acceptable levels. The most sophisticated of these programs is CLOUD(5), which takes into consideration the complete three-dimensional distribution of a group of radionuclides and their daughter products with virtually no simplifying assumptions. It also has proven difficult in practical application, owing to its complexity. 4.6.1 RADOS Program A simplified approach has been prepared by Arnett(6) at Savannah River, containing some reduction in the modeling complexity but retaining flexibility. This system, programmed by R. Cooper and known as RADOS, assumes that a cloud (or plume) is described by a double Gaussian distribution in the crosswind (y) direction and vertical (z) direction, but that these exist unchanged upwind and downwind of the position of the receptor. This assumption becomes valid after a few hundred meters of plume travel. Before this, the plume is more accurately represented by a line source. As many as 20 radionuclides and their daughter products, which are grouped into four energy classes, can be handled by RADOS in one calculation. In application to the Nine Mile Point elevation, it has been assumed that the terrain downwind is flat. This point has been NMP Unit 1 UFSAR Section XVII XVII-16 Rev. 25, October 2017 discussed in Section 4.3.1(a) in relation to concentrations, but is reconsidered here in terms of the dose rates. For the better dispersion conditions (Classes I, II and III) corrections for the slight rise in terrain south of the site would be completely insignificant. In Class IV conditions, a small correction would be in order from 5 km onward, especially on the hilltops. This would be small (<20 percent), however, and would be applicable to regions far beyond that at which the dose rates are of critical interest. 4.6.2 Centerline Dose Rates The RADOS program has been used to establish a series of centerline dose rates (directly under the cloud center as it passes downwind) for the distribution of radionuclides listed in Table XVII-27, together with their daughter products, where applicable. These dose rates are shown in Figures XVII-47 through XVII-53, where it appears that a unidirectional emission of 1 Ci/sec under almost any conditions will produce dose rates of approximately 5 R/yr at 1 km, although a Class IV condition will produce a rate 30 times that of a Class I at 50 km. The use of R/yr as an ordinate is a bit misleading, since these plots apply only to periods during which conditions and wind directions are stationary. However, comparison with later results is made easier by its use. 4.6.3 Sector Dose Rates The RADOS centerline gamma dose rates are not convertible to sector doses by the same procedures applied to concentrations (page A-11 of the PHSR). This is because the gamma dose rate is not necessarily similar to the centerline concentration. As an extreme example, assume a very confined plume high above the ground as shown on Figure XVII-54. Such a case might be observed close to the stack in inversion conditions. The concentration distribution in the y-direction is represented by the narrow and peaked upper dotted curve. The corresponding gamma dose rate distribution on the ground, represented by the lower dotted curve, is relatively broad and flat, quite unlike the concentration distribution. At greater distances where the cloud becomes quite broad, the relative shapes of the concentration and radiation distributions would become very similar, reflecting the fact that the radiation at any point is derived essentially from a uniform concentration distribution. In this case it would be perfectly NMP Unit 1 UFSAR Section XVII XVII-17 Rev. 25, October 2017 acceptable to use the PHSR procedure to go from centerline to sector dose rates. The problem is thus to determine at what distance from the source the radiation dose rate begins to be directly proportional to that of the concentration. This is easily assessed by plotting on a log-log scale the radiation dose rate as a function of cloud width (expressed by the crosswind standard deviation, y). Whenever, for a given distance and turbulence condition, the dose rate begins to vary linearly with y, we may safely assume that the radiation is essentially being received from a uniform concentration. On Figure XVII-55 such a plot is shown for the Class IV condition at 1 km. The curved, solid line representing the gamma dose rate approaches a straight line at a y value of approximately 200 m. Such a value of y, however, is not typical of a Class IV plume at 1 km. On the contrary, a proper value would be 42 m, so we know that the ground-level dose rate pattern is not similar to the concentration distribution in our region of interest. Application of the PHSR centerline-sector conversion system would give unrealistically low values. Figure XVII-55, however, provides a means of determining a correction factor for the conversion. If we extend the straight portion of the dose rate curve upward to the left, we can estimate a fictitious centerline dose rate that would be observed if the dose rate and concentration curves were really similar. In the example shown, this fictitious value is 1.36 R/yr at the typical y of 42 m, compared to the actual value of 0.7l. Since the effect of long-term dispersion within a sector is to spread the cloud laterally (increase y), we may use this fictitious value as a basis for conversion to the radial dose rates and the sector values. The correction factor used to obtain the fictitious centerline dose rate approaches unity as the cloud grows wider, as is reflected in the complete set of correction factors given in Table XVII-28. These centerline dose rates were then converted to the average annual sector dose rates of Table XVII-29 by application of wind direction and turbulence statistics. This conversion technique is believed to be conservative in the sense of overestimating the dose rates close to the source, primarily because the basic assumptions of RADOS are conservative. The plume is presumed to extend infinitely upwind NMP Unit 1 UFSAR Section XVII XVII-18 Rev. 25, October 2017 and downwind from the point of estimation; this obviously cannot be true upwind of the stack itself. 4.7 Concentrations From a Major Steam Line Break The behavior of hot gases released from a major steam line break is a challenging problem since it is usually difficult to specify the initial condition of the cloud in the atmosphere. There is little justification for assuming that an effluent which could escape from various vents and fissures in a building can properly be described as a single, cohesive cloud. Therefore, no rise of the cloud resulting from buoyancy is assumed. The treatment described on page A-17 of the PHSR is still considered appropriate. 5.0 Conclusions There is no evidence of any important difference between meteorological conditions during 1964 and those reported in the PHSR. There are variations, of course, such as slightly different wind direction patterns, but none have been found which have a significant bearing on either the routine dispersion patterns or the estimates of limiting dispersion conditions. The meteorological estimates of accident situations remain very much as they were. Thus, the additional data accumulation and analysis have served to reinforce the favorable conclusions reached in the PHSR. B. LIMNOLOGY 1.0 Introduction In 1963 an extensive limnological program was carried out in the region of Nine Mile Point, Lake Ontario, off the site of the Nine Mile Point Nuclear Station (Figure XVII-56). More than 80 cruises were made to gather information about the waters of the lake. The data obtained and conclusions drawn were presented in the PHSR, Volume II, Appendix B. The material presented in that document was concerned with the following. 1. Defining the offshore currents at Nine Mile Point. 2. Correlating these currents with various wind regimes which might then be used as indicators of water current patterns. NMP Unit 1 UFSAR Section XVII XVII-19 Rev. 25, October 2017 3. Determining dilution factors applicable to effluent water released at the Nine Mile Point site. In 1964 the limnological studies were continued mainly to provide more precise information on dilution of liquid Station effluent at selected areas. In addition, a study was conducted to determine the nature of aquatic life in the vicinity of the site. Extensive field work was conducted to assure satisfactory conclusions for each of these objectives. This report includes a brief description of this work, as well as a summary of results. 2.0 Summary Report of Cruises The results of 35 cruises are used in this report. The areas worked in detail were off Oswego and Nine Mile Point. In the Oswego area three types of work were carried out. 1. Bathythermograph (BT) temperature profile data were collected along transects out into the lake. This was done in conjunction with similar transects off Nine Mile Point. The purpose of these profiles was to establish the instantaneous height of the isothermal layers at each area. 2. Intensive BT profiles were taken across the City of Oswego water intake. 3. Current studies were made using chloride concentration, temperature and current direction data. This work was done only during periods when there was a definite westward moving current. Off the Nine Mile Point site the work was concentrated on the inshore and covered the following. 1. BT profiles were coordinated with similar data off the Oswego area. 2. Chloride, temperature and current direction data were taken to determine the inshore current directions, particularly as to possible beaching. 3. One cruise involved the release of dye at the discharge area to estimate diffusion and mixing activity. NMP Unit 1 UFSAR Section XVII XVII-20 Rev. 25, October 2017 4. Samples were taken throughout the summer and fall at three prescribed lake stations to determine the distribution of plankton and possible presence of fish eggs and larvae in the area. 5. One cruise was devoted to exploration of the lake bottom. 3.0 Dilution of Station Effluent in Selected Areas In the PHSR, calculation of dilution factors for liquid Station effluent was based on surface dilution data as found for the Oswego River and modified by Lake Ontario current data. Part of the 1964 program was directed toward developing dilution factors at selected areas. For this study, considerations such as dilution of the effluent from a submerged discharge, surface drawdown into a submerged intake, tilting of isothermal planes in the lake, and velocity of local lake currents also enter into the dilution calculations. 3.1 Dilution of Effluent at the Lake Surface Above the Discharge With the discharge structure at the lake bottom, dilution will take place because of shear stresses between the effluent jet and the lake water, resulting ultimately in turbulent diffusion throughout the entire effluent water column. Calculation of the dilution upon effluent rise to the surface is based on work by Rawn et al, discussing the dilution factor in a paper entitled "Diffusers for Disposal of Sewage in Sea Water."(7) Such diffusers are subsurface discharge structures similar in principle to that for this Station. Density differences between the fresh water of the sewage and the sea water referred to in Rawn's work are paralleled by temperature (hence density) differences in the discharge of the coolant water and the lake water at Nine Mile Point. Each of the six vertical ports of the discharge structure is considered as an independent discharge opening. The dilution (So) of the effluent on rising from the lake bottom is a function of two variables (the ratio of depth to port diameter, and the Froude Number). NMP Unit 1 UFSAR Section XVII XVII-21 Rev. 25, October 2017 Where: yo = depth from center of discharge port to surface (ft) D = diameter of the discharge port (ft) F = Froude Number The Froude Number for a jet discharging horizontally into water is: Where: V = velocity of jet (ft/sec) g' = (ft/sec2) g = acceleration of gravity = 32.2 ft/sec2 s = specific gravity of effluent s = difference between specific gravities of the effluent and lake water Prediction of the dilution factor is based upon an extrapolation of the graph in Figure 8 in Rawn, and presented here on Figure XVII-57. In this solution the number on the ordinate is the ratio yo/D. Using the average annual depth of 14.75 ft to the center of one port of 5.72-ft equivalent diameter will give a number of approximately 2.6 for this ratio. For the number on the abcissa, two calculations of F (the Froude Number) were made since the density between waters of 14C difference will vary considerably, depending upon whether this difference is in cold or warm water. Thus, one calculation is for water between 4C and 18C, and the other for 24C and 38C. This 14C difference is approximately equal to the expected 25°F temperature rise of the water used for cooling in the Station condensers. For 4C to 18C difference: NMP Unit 1 UFSAR Section XVII XVII-22 Rev. 25, October 2017 For 24C to 38C difference: On Figure XVII-57, with yo/D = 2.6 and F ranging from 4.35 to 7.74, an axial dilution factor of between 2.5X and 3X at the lake surface is obtained. This dilution will be partially dependent upon a continuous supply of lake water to the area of the rising effluent plumes in the amount required for the calculated dilution. If, for example, a dilution factor of 2X is to be achieved, it has been calculated that a minimum lake current on the order of .07 mph will assure a sufficient supply of water for dilution. Likewise, a current of .14 mph will assure a sufficient supply of water if a dilution factor of 3X is to be achieved. Current velocities at the discharge site calculated from current meter data and shown on Figure XVII-58 were over 0.7 mph 66 percent of the time and over .14 mph 54 percent of the time. Lake water temperature in excess of 24C will occur less than 6 percent of the time. On this basis sufficient lake water will always be supplied to the discharge area to assure an average annual effluent dilution of at least 2X at the surface above the discharge structure. Even when no natural lake current exists, active mixing may be expected. The flow of the effluent from the discharge ports will result in mixing eddies at the lower periphery and sides of the plumes. These will entrain lake water which will be carried along with the effluent outward and upward from each of the six ports. As a result, cooler water must be drawn from the bottom of the lake toward the discharge structure to replace that water captured by the effluent. 3.2 Dilution of Effluent at the Site Boundaries 3.2.1 General NMP Unit 1 UFSAR Section XVII XVII-23 Rev. 25, October 2017 The dilution factors at the borders of the Nine Mile Point site were derived primarily from local current and temperature data. The results are summarized below. Dye Study On one of the cruises a gallon of dye was released at the surface above the location planned for the discharge structure. By the time the dye patch reached the eastern border of the Nine Mile Point site, the dye had spread over a surface area approximately 400 ft by 2,800 ft, covering more than 1,000,000 sq ft. There can be no direct correlation between the dispersion of the dye and an effluent release of about 600 cfs. However, the results indicate that in the shallow inshore area, there is a very strong mixing movement due to active upwelling and turbulence arising from flow over the rough, shallow, rocky bottom. Thus, under normal conditions of eastward moving currents, the near shore effluent will be completely mixed with the lake water available for mixing. Studies of Currents in the Station Area The permanent current meter data for the period between November 6, 1963, and February 26, 1964 (recorded at the 35-ft water depth off the Station shore), were analyzed and used in estimating expected current velocities and dilution factors. Comparison of the wind data for this 3-mo period with cumulative data for 1963 and 1964 showed that percent distribution of winds by quadrants and average wind velocities, respectively, were almost identical. Periods of calm for the 3-mo period, however, were lower than in the 2-yr distribution by about 3.2 percent. Eastward currents of low velocity would, therefore, occur 1.6 percent more than indicated since approximately half of the wind-generated currents are eastward. Otherwise, this analysis may be considered to be within expected yearly variation limits. In general, the localized currents off the Station site were consistent with the more widespread patterns recorded in the 1963 study and reported in the PHSR. Beaching of surface water occurred more frequently in the 1964 data but coincided with the wind pattern in the same manner as established in the 1963 study. During the cruises it was observed that the top surface water responded almost immediately to wind stress, changing direction with each shift of the wind. The layer of water down to NMP Unit 1 UFSAR Section XVII XVII-24 Rev. 25, October 2017 approximately 2 ft, as indicated by surface drogues, also responded very quickly to wind stress. The tendency of water at the 5-ft depth, however, was to continue flowing in the established current pattern. As many as five different current directions and velocities were plotted throughout a 50-ft water column. Normally these currents were in the same general direction, but in some cases were completely opposite. Subsurface currents were often of greater velocity than surface currents. Even though the water closest to the surface may respond to wind stress and direction, complete mixing may be expected several feet in depth. More rapid mixing will occur in water 30 ft or less in depth over rough, rocky bottom. In the open lake, wave activity and upwelling are the principal forces of mixing. Dilution Factor as Related to Current Velocities at Nine Mile Point Dilution is directly related to current velocities which will supply varying amounts of water for dilution. Expected percent of time of various magnitudes of dilution were calculated using current velocity data recorded by the permanent current meter anchored offshore from the Station. In making this calculation, the recorded data were modified to give approximate surface current values. The dilution calculation was concerned only with nearshore water where the current is generally parallel to shore. It was assumed that the water available for mixing in this area was that portion between the discharge structure and the shore, and that only half the effluent water need be considered; the other half of the effluent mixes with water lakeward of the discharge structure. 3.2.2 Dilution of Effluent at the Eastern Site Boundary The wind data indicate that the effluent water will be moved eastward approximately 50 percent of the time, following the same general current patterns diagrammed in the PHSR. In addition, there are a number of associated factors which will markedly affect actual dilution at the border of the Station site. Associated Factors Affecting Dilution 1. Wind direction will greatly influence actual dilution. For more than half the time when the current is moving NMP Unit 1 UFSAR Section XVII XVII-25 Rev. 25, October 2017 eastward, the wind is blowing offshore causing upwelling along the shore with increased dilution. 2. The eastward current pattern at Nine Mile Point is such that the main current flow parallels the shore between Oswego and the western extremity of Nine Mile Point, then flows out into the lake. In only five of the 82 cruises in 1963 were eastward currents plotted that actually paralleled the shore beyond the site. In all other cases of eastward currents, there was a divergence of the current from the shore and subsequent upwelling. This divergence will occur even with winds from the northwest when surface waters down to 5 ft may be carried shoreward. 3. With northwest winds surface water may be beached, but again wind and current data indicate that northwest winds are consistently of higher velocity and are associated in time with stronger currents and high mixing conditions due to onshore wave activity. 4. Periods of surface calm would be expected to produce periods of lowest dilution, yet calms are most often associated with southerly winds, periods of current reversals in the immediate vicinity of the discharge structure and, at times, with residual currents which flow at velocities of up to .4 mph. During periods of surface calm there will be local currents induced by the flow of the effluent water from the underwater discharge structure. The resultant mixing produced by this flow was discussed in Section 3.1. Dilution Calculation Based on Water Availability Table XVII-30 represents a calculation of dilution based upon the amount of water available due to current velocity in the immediate vicinity of the discharge structure. It assumes complete mixing in the shallow water area between the discharge structure and shore due to a combination of the active mixing produced by the discharge of the effluent at the bottom, and turbulence in the shallow water as it flows over the rough, rocky bottom. The tabulated dilution factors do not include credit for water drawn from the lake by the effluent plumes or the added dilution potential available from upwelling. Dilution Factor Corrections for Low-Current Velocity NMP Unit 1 UFSAR Section XVII XVII-26 Rev. 25, October 2017 With eastward movement of the water past Nine Mile Point about 48 percent of the year, currents of less than .07 mph will be expected to occur about 16 percent of the time. This velocity will result in less than 2X total dilution if no credit is taken for upwelling or the current induced by the effluent plumes. During these periods, several factors may bring about an increase in the dilution factor at the eastern border of the Station site as a result of the modification of the time factor listed in Table XVII-30. 1. Any wind south of west will tend to move surface water away from the Nine Mile Point shore, produce upwelling and increase dilution. Once an eastward current along the Oswego-Nine Mile Point shore is established, this current will also create a similar upwelling and dilution. From studies reported in the PHSR, the surface water will be carried away from shore about half the time there is an eastward current. Significantly, this includes some 15 percent of the time (of a total of 63 percent) when winds are less than 10 mph. Under these conditions, dilutions greater than 2X will occur at the eastern border of the Station even with current velocities less than .07 mph. The 16 percent of the time when conditions would normally result in less than 2X initial dilution will, therefore, be reduced to 8 percent. 2. If an eastward current falls below .07 mph with possible dilutions of less than 2X, and this period of low velocity is less than 14 hr before a current reversal, then the effluent of less than 2X dilution will not reach the eastern border of the Station. Instead, this low dilution water will be carried westward and out into the open lake from the northwest bend of the Nine Mile Point promontory. Permanent current records show that when current velocities fall below .07 mph, about 50 percent of this time is within the 14-hr time limit before reversal. This would reduce by half again the probability of effluent of less than 2X dilution from reaching the eastern border of the site. Total time of this low dilution at the eastern border would thus be reduced to 4 percent of the year. With current reversals from west to east, water without effluent would flow along the Oswego-Nine Mile Point shore and separate the returning diluted NMP Unit 1 UFSAR Section XVII XVII-27 Rev. 25, October 2017 effluent water from the Station effluent discharged after the current reversal. 3. With winds north of west, particularly strong winds, effluent will be driven against the Nine Mile Point shore. Although it may not produce a noticeable current near the bottom, as indicated by the permanent current meter data, such winds would bring water to the shore. This would result in sinking of the water along the shore, mixing by wave activity, and dilution of the effluent. Wind data indicate that winds from the northwest quadrant blow about 15 percent of the time, or about one-third of the time when eastward currents are expected. Such winds, if sustained, will produce a reversal of the eastward current. Increased dilution due to such winds will reduce to less than 3 percent the time when dilutions of less than 2X will reach the eastern border of the Station site. Calculation of Average Annual Dilution Factors at Eastern Border of Station Site and 4,000 Ft East of Border From Table XVII-30, the average dilution factor of the effluent being carried toward the eastern border of the Station site is 6X. Since eastward currents occur during only half the total time, there is an annual dilution factor due to current velocity alone of twice this amount, or 12X. From the discussion in the last section, water of less than 2X dilution will reach the eastern border of the Station site less than 3 percent of the time. Also, for half the time of eastward movement of the water, the strong current along the Oswego-Nine Mile Point shore will increase the dilution of the effluent moving eastward by a factor of at least two. If these factors are considered, then the annual dilution factor at the eastern border of the Station site will be in the 20X to 24X range. Since the northeast bend of the Nine Mile Point promontory is almost a mile eastward of the eastern border of the Station site, the annual dilution factor of 21X, as given on page B-37 of the PHSR, is conservative. A reexamination of current patterns along the northern shore of the Nine Mile Point promontory shows that this current normally moves out into the open lake where the shore turns southward. NMP Unit 1 UFSAR Section XVII XVII-28 Rev. 25, October 2017 Dilution factors eastward from here along the shore are, thus, in the 40X or greater range, or equivalent to that predicted for Mexico Point (as described on page B-38 of the PHSR). 3.2.3 Dilution of Effluent West of the Station Site In the 1963 limnological study, efforts were directed mainly to delineate the currents in the eastern end of the lake. Of lesser significance at that time was the acquisition of data on those factors which would affect dilution rates to the west of the Station site. In the 1964 studies considerable time was spent on such factors as tilting of the isothermic planes in the eastern end of the lake and westward movement of water. Also, an examination of water movement in the immediate area of the City of Oswego intake proved to be of value in dilution factor calculations. The various individual factors are discussed below. Also included are the factors which directly affect the dilution of the effluent in the City of Oswego intake. Effluent moving westward will tend to be carried out into the open lake more than half the time. If it is beached, complete mixing will have taken place and immediate dilution factors equal to that of water moving eastward may be expected. On this basis an annual dilution factor of greater than 40X is predicted for the western boundary of the Station site. 3.3 Dilution of Effluent at the City of Oswego Intake The nearest public drinking water supply drawn from Lake Ontario is located near Oswego, approximately 8 mi southwest of the Nine Mile Point Nuclear Station. The next closest domestic water supply inlets are over 35 mi distant at Sodus Point and at Sackets Harbor. The location, depth and rate of pumping for the City of Oswego water intake are important aspects in making any estimate of dilution factors for Nine Mile Point Nuclear Station effluent that enters the intake. The intake is about 6,000 ft from shore and 8,200 ft west of the Oswego Harbor lighthouse in about 58 ft of water. The intake structure rises several feet from the bottom and is open at the top, drawing water in through a protective heavy wooden grid. The opening is about 16-ft wide. In June of 1967, the Onondaga County Water District expects to complete a water treatment plant which will use the same water intake tunnel as the City of Oswego. The rate of pumping is NMP Unit 1 UFSAR Section XVII XVII-29 Rev. 25, October 2017 expected to be 32 mgd, comprised of peak pumping rates of up to 12 mgd for the City of Oswego, and 20 mgd for the Onondaga County Water District. Design capacity of the water treatment plant is 36 mgd with a possible ultimate capacity of 72 mgd. 3.3.1 Tilting of the Isothermal Planes and Subsequent Dilution BT soundings were made in transects from the shore off Nine Mile Point and Oswego. Sets of soundings were always made on the same day to establish the slopes of the isothermal planes between these two points. These experiments were conducted as a result of findings from the 1963 limnological program which showed that with changes of wind direction and/or speed, isothermic planes tilted to the east or west. With the heated effluent water spreading out over the surface of the lake in a 2-ft layer, reversal of the tilt of the isothermal layers from east to west will result in upwelling beneath the effluent and subsequent dilution. Since rising of the isothermal planes in the Nine Mile Point area was found to be associated with westward moving currents, this tilting is an important factor in diluting effluent in the surface waters.* Depth differences between the same isothermal plane at Oswego and Nine Mile Point were found to range between 7 ft and 27 ft, with an average difference of about 14 ft. With a reversal of an eastward flow, isothermal planes will slope in the opposite direction; an average annual dilution factor of at least 7X may be predicted with confidence. 3.3.2 Dilution as a Function of Current Velocity Several cruises, using the BT, established temperature profiles about the City of Oswego intake along the direction of the current. At least four BT soundings were taken in each case at about 100-ft intervals. The plotting of one of these profiles is shown in Figure XVII-59. Knowing the width of the intake, amount of depression of the isotherms, rate of flow of the current, and rate of pumping, the following conclusions were drawn from the study: Note, however, that tilting is merely an indication of upwelling at the eastern end of the lake, and continuing easterly winds will result in continued upwelling and dilution. NMP Unit 1 UFSAR Section XVII XVII-30 Rev. 25, October 2017 1. In a down-current direction, water is normally drawn from an area almost directly over the intake structure, an area believed to be parabolic in shape. 2. The height of this parabola is determined by the velocity of the lake current which will bring an amount of water over the intake equal to the pumping rate. Some water will probably always be drawn in from the sides of this parabola. 3. With lower velocities, as the parabolic area of drawdown becomes higher and higher, greater and greater percentages of water will be drawn from the sides of this parabola. It can be calculated, under completely isothermal conditions, that current flow must be about .057 mph on the average throughout the water column (at 32 mgd) to supply sufficient water to the parabolic area above the intake to prevent surface water drawdown. This average current is equal to a flow of about .038 mph at a depth of 55 ft. At the ultimate capacity of 72 mgd, the corresponding 55-ft current flow required is about .085 mph. From data recorded by a permanent current meter in 55 ft of water at Nine Mile Point, it can be shown that currents of less than .038 mph occur about 7 percent of the time, while currents of less than .085 mph occur about 15 percent of the time. The possibility of surface drawdown for 32 mgd under isothermal conditions would be about 1:14; a minimum annual dilution factor of 14X would thus exist due to current velocity. At 72 mgd, the possibility of surface drawdown under similar conditions would be about 1:7, yielding a minimum annual dilution factor of 7X. It is anticipated that at a 72 mgd pumping rate, a modification of the intake structure would be required to prevent any vortex activity, and surface drawdown would be eliminated. Even under least favorable conditions (zero current), only a small amount of surface water will be drawn down to the intake. As a conservative figure, one part in 100 has been estimated. Thus, there is a total annual dilution factor of 1400X resulting from various aspects of current velocity (at 32 mgd). 3.3.3 Percent of Time Effluent Will be Carried to the Oswego Area NMP Unit 1 UFSAR Section XVII XVII-31 Rev. 25, October 2017 After considering the wind data (direction, variability, duration and speed), it was found that the combination of these factors that would bring effluent water over the intake would occur only 4 percent of the time. This would mean that an annual dilution factor of 25X may be credited to wind activity. 3.3.4 Mixing With Distance On page B-40 of the PHSR, it was calculated that effluent water reaching the Oswego area would mix with distance traveled and have a dilution of 20X. 3.3.5 Oswego River Water as a Buffer to Prevent Effluent From Passing Over the Intake Data on the dispersion of the Oswego River over the lake surface, when lake currents were flowing westward, were gathered on two cruises in each year (1963 and 1964). In only one of these cases did the Oswego River water fail to cover the water intake. Insufficient data exist to suggest more than a possibility of 3X dilution due to this effect. This figure, however, was not used in the final dilution calculation. 3.3.6 Summary of Annual Dilution Factors for the City of Oswego Intake Dilution resulting from tilting of isothermal planes (i.e., upwelling) 7X Dilution resulting from water velocity over the intake 14X Dilution resulting from mixing on drawdown 100X Dilution resulting from wind direction and other wind variables 25X Dilution resulting from mixing with distance 20X Total calculated annual dilution factor 4,900,000X Minimum instantaneous dilution would be a function of mixing on drawdown and distance. Thus, the probable minimum instantaneous dilution would be on the order of 2,000X. 3.4 Dilution of Effluent at the Nine Mile Point Intake NMP Unit 1 UFSAR Section XVII XVII-32 Rev. 25, October 2017 The antivortex intake structure and the thermocline between the heated discharge and the lake water supply assure that significant recirculation will not occur between the discharge and intake structures. Therefore, recirculation between these two structures has not been further considered in this report. 3.5 Summary of Dilution in the Nine Mile Point Area 1. Analysis of local current velocities near the discharge as a factor in dilution makes it possible to construct a table of probable dilution and percent of the time of such dilutions (Table XVII-30). 2. Based on current velocity data alone, dilutions of less than 2X within the Station site nearshore area may occur about 16 percent of the time. Upwellings will further reduce this time. 3. Factors such as wind direction, current reversals and strong onshore winds may be expected to result in an annual dilution factor of at least 20X at the eastern border of the Station site. 4. Since the westward currents at the site boundary follow a pattern of lakeward movement similar to the eastward currents beyond the promontory, an annual dilution factor of greater than 40X is predictable for the western boundary of the Station site. 5. The annual dilution factor for the Station effluent at the City of Oswego water intake will be at least 4,900,000X. 4.0 Preliminary Study of Lake Biota Off Nine Mile Point This phase of the study was directed toward acquiring a background of information on the aquatic organisms in the Nine Mile Point area. 4.1 Biological Studies 4.1.1 Plankton Study Samples of plankton were gathered from July through October of 1964. Species distribution and fluctuation in numbers were then determined for the major zoo-plankton components of the samples. It was found that, in general, concentrations of plankton 2 mi NMP Unit 1 UFSAR Section XVII XVII-33 Rev. 25, October 2017 offshore are higher than those close to shore. Individual sample concentrations followed the expected late spring and fall plankton "blooms." As much as ten times the number of plankters occur in the late spring as in midsummer, with the fall bloom increasing to only five times the midsummer concentration. Highest estimated concentrations of surface plankton were 75 crustaceans and rotifers per cu ft. Significantly, no fish eggs or fry were found in the plankton samples. 4.1.2 Bottom Study The lake bottom was examined by diving in the area off the Station site, in Mexico Bay and Near Oswego. In the Nine Mile Point area, the bottom is entirely rocky, gradually sloping off into deeper water by a series of steps as each bedding plane of the rock is eroded shoreward. For the most part, there are no loose rocks on the bottom until the 35- to 40-ft depth is reached; there only rectangular boulders (2 ft by 3 ft by 4 ft) were common. Down to a depth of 10 to 12 ft, the bottom is made up of smooth rock, supporting only a low mat of plant growth. At about the 15-ft depth, each of the bedding planes is broken into a series of rock slabs, the thickness of the bedding plane (6 in to 12 in) and 4 to 5 ft in width. In many cases the crack or joint between these slabs is moderately eroded and partially filled with rock debris. It was in this debris that the only bottom animals in the area were found. A few clams, snails, gammarids, insect larval cases, sponges, and occasionally cottid fish were found. A maximum of two such cracks eroded along a length of 3 to 4 ft were found in a 10-ft square. Unbroken bedding planes were found quite regularly. In an entire day of diving in the area (about 5 hr underwater), only one bass, of medium size, and one eel were seen. No other game fish were observed. Since bass are normally attracted to concluded that very few game fish occupy the area. 4.2 Summary of Biological Studies The entire stretch off Nine Mile Point is one almost barren of biological growth. Exposed as it is to heavy wave activity, the bottom is periodically scoured. This area possesses not only a very limited food supply, but no underwater structures suitable for the shelter of fish. Based on observations, this area is capable of supporting less than 10 percent of the biological NMP Unit 1 UFSAR Section XVII XVII-34 Rev. 25, October 2017 organisms per unit of area found in the more open portions of Mexico Bay. 5.0 Conclusions Very extensive field work, over 120 cruises in a 2-yr period, and a thorough compilation and examination of the resulting limnological data have yielded a detailed charting of water currents and their relation to wind patterns in the Nine Mile Point area. Only at Douglas Point in Lake Huron have comparable data been accumulated in any of the Great Lakes. In general, 1964 data served to confirm the conservative nature of the PHSR dilution factor calculations. The Nine Mile Point Nuclear Station liquid effluent, being discharged off a promontory, is carried offshore into the open lake a very high percentage of the time from either bend of the promontory. In addition, the effluent is also subjected to active upwelling and sinking, important factors in mixing and dilution, because of the Station location near the end of the lake. From what is now known of the current patterns in the lake, and what may be assumed of the current patterns, the Station appears to be located at one of the optimum sites on the southern shore of Lake Ontario. C. EARTH SCIENCES 1.0 Introduction The geological, hydrological and seismological conditions of the Nine Mile Point Nuclear Power Station site and adjacent areas, located near Oswego, New York, were detailed in the PHSR, Vol. II, Appendix C. Included in the present report are: 1. A summary of results of the subsurface investigation subsequent to the PHSR. 2. Observations made during construction, including blast monitoring results during rock excavation. 3. A correlation of subsequent investigation results with those documented in the PHSR. 2.0 Additional Subsurface Studies NMP Unit 1 UFSAR Section XVII XVII-35 Rev. 25, October 2017 As pointed out, a comprehensive study of the regional and local geology was presented in the PHSR. During the original study it was found that the bedrock underlying the site is Oswego sandstone, a light gray to greenish-gray moderately hard, dense material. The rock is extensively cross-bedded, containing silty and shaly laminations and thin, discontinuous lenses. An increase of shaly laminations with depth was found. The gradational contact of the sandstone with the underlying Lorraine shales was estimated at roughly 185 ft below the ground surface in the area of the reactor building (185-ft depth corresponds to elevation 80). In order to evaluate subsurface conditions in the intake and discharge tunnel areas, four core borings were drilled along the proposed alignment to depths ranging from 100 to 138 ft beneath the lake surface. Continuous rock cores were extracted from these borings. Locations of these borings, along with tunnel profiles, are presented on Figure XVII-60, while logs of the core borings are shown on Figures XVII-61 through XVII-64. None of the four borings drilled in connection with the tunnel study penetrated to Lorraine formation. The deepest boring was drilled to elevation +106, roughly 25 ft above the assumed contact. The rock encountered in the four borings was similar in all respects to the onsite rock described in the PHSR. Cross-bedding was prevalent and shaly laminations increased with depth. Horizontal fractures were largely coincident with shale laminae and were attributed to the stress relief which resulted from coring operations. Occasional irregular and vertical fractures, probably discontinuous joints, were found in all of the borings. No overburden was encountered at the lake bottom in three of the borings. In the fourth boring, drilled approximately 1,000 ft from the shoreline, approximately 3.5 ft of sandy and gravelly material was found overlying the rock. A large slab or boulder of sandstone was encountered above this unconsolidated material. This slab may be similar to the large slabs observed near the shoreline. Based upon recommendations resulting from this investigation, the tunnels were constructed at higher elevations than had been originally planned in order to minimize the number of shale laminations encountered and the effects of residual stresses in the rock. Centerline elevations of +185 ft for the intake and 195 ft for the discharge tunnels were used in order to leave NMP Unit 1 UFSAR Section XVII XVII-36 Rev. 25, October 2017 about 20 ft of sound structural rock over the excavation and to minimize water infiltration during construction. 3.0 Construction Experience 3.1 Station Area Prior to rock blasting in the Station area, the soil overburden (10 to 12 ft of glacial till) was removed. The reactor building area was then handswept, washed and thoroughly inspected. The exposed rock surface was similar to that examined in nearby outcrops. Ripple marks, joints and glacial striae were evident. No evidence of movement along the joints was found. The major joint systems were found to be in accord with those trends reported in the PHSR. Upon commencement of blasting activities in the reactor building area, a blast monitoring program was initiated. This monitoring was performed to assist in limiting the size of explosive charges so as to preclude damage to adjacent rock, structures and properties in the vicinity. Approximately 100 dynamite blasts were monitored using a three-component seismograph, with results recorded on photographic paper. The seismograph was located at distances varying from 25 to 750 ft from the blasts. Dynamite loads ranged from 5 to 1,250 lb. Following each blast, the condition of the rock adjacent to the affected area was inspected. In many cases cracks were evident. Some of these cracks followed joints; in other cases they were more irregular. 3.2 Intake and Discharge Tunnels An inspection of the onshore shaft and the tunnels was made on November 4 and 5, 1965. At that time the discharge tunnel had been excavated to a point about 290 ft from the shaft. Excavation of the intake tunnel had not progressed as far. Examination of the exposed rock revealed conditions consistent with those encountered during the previous studies. Rock conditions in the tunnels differ. The discharge tunnel (elevation 195) is in thicker-bedded and less shaly rock than the intake tunnel (elevation 185). No zones of defective rock were found and no weathered rock was evident; the rock in both tunnels is sound. Water infiltration was practically nonexistent, being limited to scattered areas which were only damp. The roofs of both tunnels appeared to be structurally sound. NMP Unit 1 UFSAR Section XVII XVII-37 Rev. 25, October 2017 The actual conditions found in the tunnel excavations are in agreement with those encountered in Boring CB-1, drilled during the subsurface investigation. Based upon this agreement, along with knowledge gained from previous borings, it was anticipated that the percentage of shaly strata might increase as the discharge tunnel approached its terminus approximately 550 ft from the pump house shaft. Some decrease in shaly laminations was anticipated as excavation of the intake tunnel progressed. It was estimated that this decrease would be gradual and would be noticeable to roughly 800 ft from the pump house shaft. Although an increase in the possibility of water seepage was anticipated with further extension of the tunnels out under the lake, no inordinate problems were encountered during excavation of either tunnel. In February 1965, a decision was made to extend the intake tunnel farther into the lake than was originally planned. This action was taken simply to terminate the tunnel in that depth of water specified in the original design criteria. An investigation of the lake bottom subsequently was made in the vicinity of the lake inlet along the alignment of the tunnel. The bottom of the lake was found to be of variable character. In some areas, flat or gently sloping rock is exposed. Occasional, relatively large, flat slabs of rock were found. In other areas, loose sand, gravel, cobbles and boulders cover the rock. Some of this loose material takes the form of subdued ridges roughly parallel to the shoreline. Only a small area was examined, but surface conditions were found to vary within short distances. Based on this examination, it was decided to extend the intake tunnel 90 ft farther from shore than originally planned in order to take advantage of more favorable lake bottom conditions. Slight water seepage was encountered in the intake tunnel when excavation had progressed to a point approximately 800 ft from the shaft. The leakage was effectively arrested by conventional grouting procedures from within the tunnel. 4.0 Correlation With Previous Studies 4.1 General Subsequent investigations have revealed no significant discrepancies with conditions reported in the PHSR. Inspection of rock exposed during excavations for the reactor and the tunnels has allowed a more detailed study of subsurface conditions. Cross-bedding and shaly laminations have been examined in detail. The lenticular nature of the rock actually was seen to exist, whereas it could only be referred from NMP Unit 1 UFSAR Section XVII XVII-38 Rev. 25, October 2017 previous borings. The predicted soundness of the rock has been substantiated by its behavior during blasting operations. Attenuation curves prepared from blast monitoring data, shown on Figure XVII-65, indicate that rock underlying the site is quite sound. Comparison with empirical data collected by the United States Bureau of Mines (USBM) revealed that behavior of the Oswego sandstone under dynamic loading is comparable to that of sound rock in other areas. 4.2 Geological Conditions Only Oswego sandstone has been encountered during construction of the various facilities. None of the borings or excavations penetrated the Lorraine sediments. Inspection of the reactor excavation revealed no open joints below the upper 10 to 20 ft of rock. The small amount of groundwater seepage that occurred emanated from the upper strata. No significant open joints have been encountered in the excavation for the tunnels. It appears that the conclusion stated in the PHSR regarding a lack of open joints at depths more than 20 ft below the rock surface has been substantiated. The rock has been observed to be sound, cross-bedded, and relatively thinly laminated. Shaly strata appear to increase with depth. No discontinuities, folds, or faults have been found. The rock appears to conform in all respects with the description presented in the PHSR. 4.3 Hydrological Conditions The flow of groundwater appears to be consistent with conclusions stated in the PHSR. No significant flow of water was encountered at depths greater than 10 to 20 ft below the surface of the rock. Horizontal bedding limits, vertical flow of water through the rock itself, and the cross-bedded nature of the rock precludes any horizontal flow over any appreciable distance. The rock appears to be practically impermeable at depths sufficient to prevent relief of stresses and consequent open joints. During operation of Unit 2, the groundwater near Unit 1 building structures is influenced by operation of the Unit 2 reactor building dewatering system. Details of the Unit 2 dewatering system are provided in the Unit 2 Updated Safety Analysis Report (USAR). NMP Unit 1 UFSAR Section XVII XVII-39 Rev. 25, October 2017 4.4 Seismological Conditions No faults have been encountered during the more recent phases of investigation and inspection. No unusual conditions were encountered which would necessitate a revision in the conclusions reached in the PHSR. 4.5 Conclusion The additional investigations and recent field inspections have served to reinforce conclusions reached in the PHSR. Geological, hydrological and seismological conditions have been found to be consistent with those described in the PHSR and are favorable for a nuclear plant site. D. REFERENCES 1. M. E. Smith and I. A. Singer. "An Improved Method of Estimating Concentrations and Related Phenomena from a Point Source Emission," Journal of Applied Meteorology, October 1966. 2. E. W. Hewson et al. "Topographic Influences on the Behavior of Stack Effluents," Proceedings, American Power Conference, XXIII, 1961. 3. R. E. Munn and T. L. Richards. "The Lake Breeze: A Survey of the Literature and Some Applications to the Great Lakes," University of Michigan, Great Lakes Research Division, Publication No. 11, 1964. 4. R. J. Englemann, R. W. Perkins, D. I. Hagen and W. A. Haller. "Washout Coefficients for Selected Gases and Particulates," Battelle Northwest Laboratory, BNWL-SA-657, April 1966. 5. D. S. Duncan. CLOUD - An IBM 709 Program for Computing Gamma-Ray Dose Rate from a Radioactive Cloud," NAA-SR Memo 4822, Atomics International, 1959. 6. L. M. Arnett. "Calculation of Radiation Dose from a Cloud of Radioactive Gases," Nuclear Applications, April 1967. 7. A. M. Rawn, F. R. Bowerman, and N. H. Brooks. Journal of the Sanitary Engineering Division, Proceedings of the NMP Unit 1 UFSAR Section XVII XVII-40 Rev. 25, October 2017 American Society of Civil Engineers, Volume 86, No. SA2: 65-106, March 1960. NMP Unit 1 UFSAR Section XVII XVII-41 Rev. 25, October 2017 TABLE XVII-1 DISPERSION AND ASSOCIATED METEOROLOGICAL PARAMETERS Trajectory Turbulence Class Sutton Equation Gaussian Equation Mean Wind Speed Wind Profile Exponent Inversion Lid Height n Cy Cz u (30 ft) (350 ft) P - (mn/2) (m) (m/sec) - (m) From Lake I .19 .56 .58 .40x.91 .41x.91 2.0 3.5 .17 1500 From Lake II .28 .50 .46 .36x.86 .33x.86 4.0 8.0 .29 1500 From Lake III .45 .45 .32 .32x.78 .22x.78 6.0 13.0 .32 500 From Lake IV .58 .44 .05 .31x.71 .06x.71 2.0 8.0 .49 200 From Land I .19 .56 .58 .40x.91 .41x.91 3.0 4.5 .17 1500 From Land II .28 .50 .46 .36x.86 .33x.86 3.0 6.0 .29 1500 From Land III .45 .45 .32 .32x.78 .22x.78 5.0 10.5 .32 500 From Land IV .58 .44 .05 .31x.71 .06x.71 1.5 5.0 .49 200 NOTE: The values of n, Cy, Cz, , and are specifically chosen to represent the upper-level conditions at approximately 350 ft. They are used to represent low-level conditions also, even though actual conditions closed to the ground would generally be affected by somewhat more favorable conditions. NMP Unit 1 UFSAR Section XVII XVII-42 Rev. 25, October 2017 TABLE XVII-2 RELATION OF SATELLITE AND NINE MILE POINT WINDS CALM 6 9 12 4 3 5 2 4 11 17 39 19 12 13 7 1 NNW 18 6 1 1 1 6 32 NW 9 1 1 2 5 11 59 32 WNW 3 3 4 10 31 86 39 4 W 1 1 1 12 104 136 34 2 WSW 2 3 37 118 32 4 1 SW 1 2 1 2 37 103 54 26 4 7 SSW 2 2 1 1 11 54 49 8 10 3 2 S 4 1 2 1 9 87 62 27 4 3 4 1 SSE 3 2 1 1 3 38 62 7 9 1 SE 2 1 4 2 2 5 72 78 48 4 5 2 1 1 2 ESE 1 1 3 5 50 129 20 16 1 4 1 E 2 4 19 41 40 6 4 2 1 ENE 3 1 7 5 7 6 13 4 4 NE 2 2 19 10 6 1 12 5 NNE 5 39 29 2 3 1 2 6 N 34 46 5 6 3 1 7 N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW CALM NOTE: Figures are in hours. NMP Unit 1 UFSAR Section XVII XVII-43 Rev. 25, October 2017 TABLE XVII-3 FREQUENCY OF OCCURRENCE OF LAPSE RATES 1963 & 1964 Temperature Difference - Frequency of Occurrence (Percent) Jan Feb Mar April May June July Aug Sept Oct Nov Dec* -15.0 to -1.1 62.9 58.6 49.7 46.4 32.3 32.9 37.1 48.2 37.6 34.8 31.2 67.6 -1.0 to -0.6 16.8 17.1 15.3 12.7 8.9 9.1 9.5 13.4 15.5 11.6 18.4 12.2 -0.5 to +0.5 11.0 13.4 15.7 16.5 19.0 16.3 17.3 13.1 16.7 14.4 28.2 10.4 0.6 to 4.9 9.1 10.0 18.0 21.6 30.2 34.5 31.6 21.8 24.5 33.7 20.3 9.3 5.0 to 35.0 0.2 0.9 1.3 2.8 9.6 7.2 4.5 3.5 5.7 5.5 1.9 0.5

• NMP Unit 1 UFSAR Section XVII XVII-44 Rev. 25, October 2017 TABLE XVII-4 RELATION BETWEEN WIND DIRECTION RANGE AND TURBULENCE CLASSES (Percent) Turbulence Class Angular Range 0-9° 10-19° 20-29° 30-39° 40-49° 50-59° 60-69° 70-79° 80°+ I <1 1 1 1 5 31 28 11 21 II 8 37 33 16 5 1 <1 <1 <1 III <1 14 25 35 20 5 1 <1 <1 IV Day 98 <1 -- -- -- -- <1 -- <1 IV Night 99 <1 <1 -- -- -- -- -- <1 Mean 41 10 12 10 6 7 6 2 4 NMP Unit 1 UFSAR Section XVII XVII-45 Rev. 25, October 2017 TABLE XVII-5 STACK CHARACTERISTICS Emission Parameters Ev - (Volume of emission at ambient temperature) 3.14 x 103 ft3/sec Vs - (Speed of emission) 55 ft/sec _ u - (Wind speed) Variable ft/sec T - (Ambient temperature) 10°C - (Temperature excess of stack gas) 30°C G - (Stability parameter) .001 & .030°C/ft Hs - (Actual stack height) 350 ft (107m) Effective Stack Heights for Pertinent Cases* Wind Speed Turbulence Class (m/sec) I, II, III IV 3.5 205 m 4.5 168 5.5 144 6.0 146 8.0 133 128 10.5 125 13.0 119
• Computed from the Bosanquet, Carey & Halton (1950) equation.

NMP Unit 1 UFSAR Section XVII XVII-46 Rev. 25, October 2017 TABLE XVII-6 DISTRIBUTION OF TURBULENCE CLASSES BY SECTORS Turbulence Class Sector I II III IV Day IV Night A 1.5 12.2 6.0 0.8 3.1 B 1.4 7.9 3.9 0.1 1.0 C 0.2 6.0 3.3 0.1 0.4 D1 0.4 16.4 5.3 1.6 0.9 D2 0.1 9.2 0.5 0.8 0.4 E 0.1 2.8 <0.1 0.4 0.2 F 0.1 6.2 0.3 1.1 0.6 G <0.1 4.0 0.1 0.3 0.5 NMP Unit 1 UFSAR Section XVII XVII-47 Rev. 25, October 2017 TABLE XVII-7 SECTOR CONCENTRATIONS 1963-64 Sector A Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 1.9 4.1 0.6 0.6 3.6 9.1 1.0 0.8 4.4 11.6 1.2 1.0 4.0 10.6 1.1 3.0 1.2 2.3 0.6 4.0 5.0 0.5 0.9 0.3 10.0 0.14 0.2 0.07 16.0 20.0 50.0 0.03 0.04 0.01 NMP Unit 1 UFSAR Section XVII XVII-48 Rev. 25, October 2017 TABLE XVII-8 SECTOR CONCENTRATIONS 1963-64 Sector B Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 3.0 4.9 0.8 0.6 5.7 10.0 1.4 0.8 6.5 12.5 1.4 1.0 5.9 11.9 1.2 3.0 1.6 4.3 0.2 4.0 5.0 0.7 2.0 0.06 10.0 0.2 0.5 <0.01 16.0 20.0 50.0 0.04 0.08 0.01 NMP Unit 1 UFSAR Section XVII XVII-49 Rev. 25, October 2017 TABLE XVII-9 SECTOR CONCENTRATIONS 1963-64 Sector C Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 3.5 6.8 0.8 0.6 7.6 15.0 2.0 0.8 10.1 20.6 2.9 1.0 9.4 18.7 2.9 3.0 3.3 5.2 1.3 4.0 5.0 1.6 3.3 (4.1)* 0.7 10.0 0.6 1.7 0.2 16.0 25.0 0.2 0.7 0.05 50.0 0.11 0.30 0.03

• Data for two different months are given for some radial distances because the highest calendar month changed with radial distances.

NMP Unit 1 UFSAR Section XVII XVII-50 Rev. 25, October 2017 TABLE XVII-10 SECTOR CONCENTRATIONS 1963-64 Sector D1 Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 5.4 9.8 1.7 0.6 13.5 24.6 4.3 0.8 12.3 22.0 3.8 1.0 10.3 18.0 3.1 3.0 2.4 3.2 0.6 4.0 5.0 1.1 1.2 (1.9)* 0.2 10.0 0.2 0.2 (0.5)* 0.04 16.0 0.1 0.13 (0.28)* 0.02 20.0 0.1 0.25 (0.40)* 0.02 50.0 0.05 0.07 (0.16)* 0.01

• Data for two different months are given for some radial distances because the highest calendar month changed with radial distances.

NMP Unit 1 UFSAR Section XVII XVII-51 Rev. 25, October 2017 TABLE XVII-11 SECTOR CONCENTRATIONS 1963-64 Sector D2 Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 2.9 6.4 1.2 0.6 7.4 16.0 3.1 0.8 6.6 14.3 2.8 1.0 5.4 11.6 2.3 3.0 1.0 2.1 0.4 4.0 0.6 1.1 (1.5)* 0.4 5.0 0.5 0.8 (1.1)* 0.3 10.0 0.1 0.2 (0.32)* 0.09 16.0 20.0 50.0 0.02 0.02 (0.06)* 0.01

• Data for two different months are given for some radial distances because the highest calendar month changed with radial distances.

NMP Unit 1 UFSAR Section XVII XVII-52 Rev. 25, October 2017 TABLE XVII-12 SECTOR CONCENTRATIONS 1963-64 Sector E Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 1.8 5.8 0.2 0.6 4.6 14.5 0.4 0.8 4.2 13.0 0.4 1.0 3.4 10.6 0.3 3.0 0.6 1.9 0.04 4.0 0.5 1.0 0.2 5.0 0.3 0.7 0.1 10.0 0.07 0.14 (0.17)* 0.02 16.0 20.0 50.0 0.01 0.02 (0.04)* <0.01

• Data for two different months are given for some radial distances because the highest calendar month changed with radial distances.

NMP Unit 1 UFSAR Section XVII XVII-53 Rev. 25, October 2017 TABLE XVII-13 SECTOR CONCENTRATIONS 1963-64 Sector F Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 2.0 6.1 1.4 0.6 5.0 15.7 3.7 0.8 4.5 14.2 3.3 1.0 3.7 11.6 2.7 3.0 0.6 2.1 0.6 4.0 0.5 1.2 0.3 5.0 0.3 0.9 0.2 10.0 0.1 0.2 0.04 16.0 20.0 50.0 0.02 0.02 (0.05)* <0.01

• Data for two different months are given for some radial distances because the highest calendar month changed with radial distances.

NMP Unit 1 UFSAR Section XVII XVII-54 Rev. 25, October 2017 TABLE XVII-14 SECTOR CONCENTRATIONS 1963-64 Sector G Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 1.50 5.8 0.30 0.6 4.00 14.20 0.70 0.8 3.5 12.8 0.6 1.0 2.9 10.5 0.5 3.0 0.5 1.9 0.1 4.0 0.4 1.0 0.15 5.0 0.25 0.7 0.06 10.0 0.07 0.1 (0.11)* 0.03 16.0 20.0 50.0 0.02 0.02 (0.05)* 0.01

• Data for two different months are given for some radial distances because the highest calendar month changed with radial distances.

NMP Unit 1 UFSAR Section XVII XVII-55 Rev. 25, October 2017 TABLE XVII-15 SECTOR CONCENTRATIONS 1963-64 Sector A Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 925 1528 393 0.6 391 648 169 0.8 218 363 94 1.0 136 226 59 3.0 14.6 22.9 6.0 4.0 5.0 5.3 8.6 1.9 10.0 1.3 2.3 0.5 16.0 20.0 50.0 0.13 0.4 <0.1 NMP Unit 1 UFSAR Section XVII XVII-56 Rev. 25, October 2017 TABLE XVII-16 SECTOR CONCENTRATIONS 1963-64 Sector B Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 805 1443 91 0.6 343 613 39 0.8 193 343 22 1.0 120 214 13 3.0 12.4 22.0 1.0 4.0 5.0 4.6 8.0 0.3 10.0 1.2 2.0 0.1 16.0 20.0 50.0 0.1 0.3 <0.1 NMP Unit 1 UFSAR Section XVII XVII-57 Rev. 25, October 2017 TABLE XVII-17 SECTOR CONCENTRATIONS 1963-64 Sector C Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 842 1601 246 0.6 374 716 108 0.8 199 404 60 1.0 133 261 38 3.0 16 31 4 4.0 5.0 7.5 13.0 1.4 10.0 3.0 4.0 0.2 16.0 20.0 50.0 0.1 0.4 <0.1 NMP Unit 1 UFSAR Section XVII XVII-58 Rev. 25, October 2017 TABLE XVII-18 SECTOR CONCENTRATIONS 1963-64 Sector D1 Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 1069 3429 80 0.6 476 1536 34 0.8 266 868 19 1.0 172 561 12 3.0 20 56 1.3 4.0 5.0 7.1 25.0 <0.1 10.0 1.9 6.0 <0.1 16.0 20.0 50.0 0.1 0.5 <0.1 NMP Unit 1 UFSAR Section XVII XVII-59 Rev. 25, October 2017 TABLE XVII-19 SECTOR CONCENTRATIONS 1963-64 Sector D2 Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 524 2625 128 0.6 233 1177 55 0.8 130 667 30 1.0 85 432 20 3.0 59 33 2 4.0 5.0 2.0 12.0 0.3 10.0 0.5 4.0 0.1 16.0 20.0 50.0 <0.1 0.3 <0.1 NMP Unit 1 UFSAR Section XVII XVII-60 Rev. 25, October 2017 TABLE XVII-20 SECTOR CONCENTRATIONS 1963-64 Sector E Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 465 1404 68 0.6 207.5 584.0 29.0 0.8 115.4 331.0 16.0 1.0 75.3 214.0 10.0 3.0 4.4 14.0 1.1 4.0 5.0 1.4 5.0 0.4 10.0 0.4 1.3 0.1 16.0 20.0 50.0 <0.1 0.1 <0.1 NMP Unit 1 UFSAR Section XVII XVII-61 Rev. 25, October 2017 TABLE XVII-21 SECTOR CONCENTRATIONS 1963-64 Sector F Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 564 1814 82 0.6 250 803 35 0.8 140 462 20 1.0 91 300 12 3.0 5.8 18.0 1.1 4.0 5.0 1.9 6.0 0.4 10.0 0.6 2.0 0.1 16.0 20.0 50.0 <0.1 0.1 <0.1 NMP Unit 1 UFSAR Section XVII XVII-62 Rev. 25, October 2017 TABLE XVII-22 SECTOR CONCENTRATIONS 1963-64 Sector G Source Height Ground Radial Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) Two-Year Average Highest Calendar Month Lowest Calendar Month 0.4 529 1837 51 0.6 236 780 21 0.8 133 443 12 1.0 87 287 7 3.0 7.4 23.0 0.8 4.0 5.0 2.5 8.5 0.3 10.0 0.7 2.2 0.1 16.0 20.0 50.0 0.1 0.3 <0.1 NMP Unit 1 UFSAR Section XVII XVII-63 Rev. 25, October 2017 TABLE XVII-23 ESTIMATES OF THE LEAST FAVORABLE 30 DAYS IN 100 YEARS (Concentrations in units/meter3 x 10-8 for a 1 unit/sec release) Synoptic Estimates Ground-Level Release Elevated Release Distance (km) Sector D Other Sectors All Sectors 0.2 25,000 15,000 40 1.0 1,500 900 70 10.0 25 15 2.5 Statistical Estimates Ground-Level Release Sectors Distance (km) A B C D E F G 0.2 10,500 7,800 7,750 15,500 7,900 12,200 9,000 1.0 630 470 465 930 473 735 540 10.0 10.5 7.8 7.8 15.5 7.9 12.2 9.0 NMP Unit 1 UFSAR Section XVII XVII-64 Rev. 25, October 2017 TABLE XVII-24 CONCENTRATIONS IN THE LEAST FAVORABLE CALENDAR MONTH 1963-64 Distance (km) Concentrations (units/meter3 x 10-8 for a 1 unit/sec release) A B C D1 D2 E F G Ground-Level Release 0.2 6,417 6,102 6,238 13,332 10,279 5,150 7,106 6,799 1.0 226 214 261 561 432 214 300 287 10.0 2.3 2.0 4.0 6.0 4.0 1.3 2.0 2.2 Elevated Release 0.2 <0.1 0.1 -- 0.1 <0.1 <0.1 <0.1 <0.1 1.0 10.6 11.9 18.7 18.0 11.6 10.6 11.6 10.5 10.0 0.2 0.5 1.7 0.5 0.3 0.2 0.2 0.1 NMP Unit 1 UFSAR Section XVII XVII-65 Rev. 25, October 2017 TABLE XVII-25 ANNUAL AVERAGE SECTOR DEPOSITION RATES (Vg = 0.5 cm/sec) Radial Distance (km) Deposition (units/m2 sec x 10-11 for a 1 unit/sec release) A B C D1 D2 E F G 0.6 18 29 39 70 38 24 25 21 1.0 21 30 48 53 27 17 18 14 3.0 6 8 18 12 5 3 3 2.5 4.0 3.1 2.5 2.5 2.0 5.0 2.6 3.6 8.2 5.7 2.5 1.5 1.5 1.3 10.0 0.7 1.0 3.1 1.0 0.5 0.4 0.5 0.4 16.0 0.5 20.0 0.5 25.0 1.0 50.0 0.15 0.21 0.57 0.25 0.10 0.05 0.10 0.10 NMP Unit 1 UFSAR Section XVII XVII-66 Rev. 25, October 2017 TABLE XVII-26 ANNUAL AVERAGE SECTOR DEPOSITION RATES (Vg = 2.5 cm/sec) Radial Distance (km) Deposition (units/m2 sec x 10-11 for a 1 unit/sec release) A B C D1 D2 E F G 0.6 90 145 195 350 190 120 125 105 1.0 105 150 240 265 135 85 90 70 3.0 30 40 90 60 25 15 15 13 4.0 16 13 13 10 5.0 13 18 41 29 13 8 8 7 10.0 3.5 5.0 16 5.0 2.5 2.0 2.5 2.0 16.0 2.5 20.0 2.5 25.0 5.0 50.0 0.75 1.05 2.85 1.25 0.50 0.25 0.50 0.50 NMP Unit 1 UFSAR Section XVII XVII-67 Rev. 25, October 2017 TABLE XVII-27 PRINCIPAL RADIONUCLIDES IN GASEOUS WASTE RELEASE* (Note: 3.0(3) = 3.0 x 103, etc.) Radionuclide Ci/sec Radionuclide Ci/sec NOBLE GAS COMPOSITION Kr-83m 2.5(4) Xe-133m 2.0(3) Kr-85m 5.8(4) Xe-135 1.8(5) Kr-87 1.6(5) Xe-135m 8.0(4) Kr-88 1.7(5) Xe-137 9.0(3) Kr-89 3.0(3) Xe-138 2.7(5) Xe-133 5.0(4) PARTICULATES (Assuming 99 Percent Filter Efficiency) Rb-88 1.2(3) Y-94 1.4(1) Rb-89 1.2(3) Cs-137 4.6(-3) Rb-90 5.8 Cs-138 2.4(3) Sr-89 4.8(-1) Cs-139 3.4(2) Sr-90 1.7(-3) Ba-139 2.4(2) Sr-91 1.7(1) Ba-140 7.2(-1) Sr-92 2.2(1) Ba-141 2.8(1) Sr-93 1.2(1) La-140 5.6(-3) Y-90 1.5(-3) La-141 4.6 Y-91 2.2(-3) La-143 3.0 Y-91m 2.2 Ce-141 1.2(-3) Y-92 2.0 Ce-143 5.8(-2) Y-93 2.6 HALOGENS Br-84 4.9(-2) I-133 3.9(-1) I-131 9.8(-2) I-134 4.9(-1) I-132 2.9(-2) I-135 4.9(-1)

• Basis - Noble gas (diffusion type mixture) release rate at stack = 1.0 Ci/sec following holdup time of 30 min.

NMP Unit 1 UFSAR Section XVII XVII-68 Rev. 25, October 2017 TABLE XVII-28 CORRECTION FACTORS TO OBTAIN ADJUSTED CENTERLINE DOSE RATES FOR SECTOR ESTIMATES Correction Factors Distance (km) Classes I and II Class III Class IV 1 1.3 1.5 1.9 3 1.1 1.2 1.6 5 1.05 1.1 1.4 10 1 1 1.2 20 1 1 1 50 1 1 1 NMP Unit 1 UFSAR Section XVII XVII-69 Rev. 25, October 2017 TABLE XVII-29 ANNUAL AVERAGE GAMMA DOSE RATES (Release Rate = 1 Ci/sec) Radial Distance (km) Sector Dose Rates (R/year) A B C D1 D2 E F G 1.0 0.26 0.40 0.64 0.78 0.40 0.25 0.27 0.22 2.0 0.12 0.18 0.30 0.36 0.17 0.10 0.13 0.11 3.0 0.08 0.11 0.19 0.21 0.09 0.06 0.07 0.06 4.0 0.047 0.078 0.134 0.134 0.056 0.034 0.045 0.034 5.0 0.032 0.048 0.085 0.084 0.034 0.028 0.034 0.028 10.0 0.011 0.016 0.052 0.029 0.017 0.011 0.009 0.014 18.0 0.013 20.0 0.004 0.006 0.030 0.011 0.004 0.003 0.002 0.003 25.0 0.027 50.0 0.002 0.002 0.004 0.003 0.001 0.001 0.001 0.001 NMP Unit 1 UFSAR Section XVII XVII-70 Rev. 25, October 2017 TABLE XVII-30 DILUTION CALCULATION FOR EASTWARD CURRENTS BASED ON WATER AVAILABILITY Current in MPH Percent of Year Dilution Factor <.07 16.1 <2X .07 - .14 5.5 2X - 3X .14 - .2 3.7 3X - 4X .2 - .5 10.1 4X - 8X .5 - 1.0 9.4 8X - 16X 1.0 - 2.0 3.0 16X - 30X Total 47.8 Average 6X ( -\ / AVERAGE WI ND ROSES FOR JANUARY '63-'64 KEY WINO SPEED -I. 10 MPH -11. 20 MPH -21.100 MPH KEY WINO SP EE D -I. 2 MPH -3. 6 MPH -7. I 0 MPH N s TOTAL WIND N s WIND < 10 M,N Pl!8URI XVY..1 UFSAR Aev. 15 t-Novemt>ar 1987) AVERAGE WIND ROSES FOR FEBRUARY 163-164 ICE Y WIND SPEED w -1. 10 MPH -----11-20 MPH 21-100 MPH KEY WIND SPEED 1. 2 MPH 3-6 MPH 7 -10 MPH N s TOTAL WIND N WIND < 10 &*M FIBOH E UFSAR Rn. 1.5 (.Nevember 1997) w KEY WIND SPEED -1. lOMPH -11. 20 MPH 100 MPH w KEY -WIND SPEED 2 MPH 6 MPH 010 MPH AVERAGE WI ND ROSES FOR MARCH 163-164 N E s TOTAL WIND N E s WIND <* 10 MPM flGURI XVll-3 UF'SAR Rev. 11(Novemtler1987) w KEY WIND SPEED -1. 10 MPH -----11. 20 MPH 21.100 MPH w KEY -WIND SPEED I. 2 MPH 3. 6 MPH 7. 10 MPH AVERAGE WIND ROSES FOR APRIL 163-164 N E s TOTAL WIND N E AGUIEXW""4 s UPSAR Rev. 15 (.November 1997) WIND < 10 MPH w KEY WIND SPEED --I. 10 MPH 11. 20 MPH -21.100 MPH w KEY -WINO $PEED -I. 2 MPH -3. 6 MPH -7. 10 MPH AVERAGE WIND ROSES FOR MAY 163-164 N E s TOTAL WIND N E .-----Sst s FIGUREXVH w IND < 10 Mp H UF&Afl R9'V. 16 CNowmlNJ 1187) w KEY WINO SPEED 10MPH -----11-20MPH 21.100 MPH w KEY -WINO SPEED !. 2 MPH 3-6 MPH 7-10 MPH AVERAGE WIND ROSES FOR JUNE '63-'64 N E s TOTAL WIND N E SSE s RGUN! xvu..e WI N D < 1 0 M P ff W8AR Rav. 15 (ilfeVfHlilbl' 1917) ( AVERAGE WIND ROSES FOR JULY '63-61.l' N w E KEY WINO SPEED 10 MPH --11-20MPH 21-100 MPH w KEY WINO SPEED -l-2 MPH 6 MPH IOMPH s TOTAL WIND N I 0 "lo E s NUMXVl ... J WIND < 10 MPH UISAll Rev. 11tN:ovember1997t ( \ \ AVERAGE WIND ROSES FOR AUGUST '63-164 KEY WINO SPEED w -1. 10 MPH -----11. 20 MPH 21.100 MPH w KEY WINO SPEED 1. 2 MPH 3. 6 MPH 7. 10 MPH N E s TOTAL WIND N E s MUM xvu .. 1 . . . . -. . WIND < 10 MPH lWSAll llev. 15 U>>ev,ember 1987)

, AVERAGE WIND ROSES FOR SEPTEMBER 163-164 KEY WINO SPEED w -I. 10 MPH -11. 20MPH -21.100 MPH w KEY WIND SPEED -1. 2 MPH -3. 6 MPH -7. I 0 MPH N E s TOTAL WIND N E SSE s RGURIXW ... WIND < 10 MPH UPW ... tlfllove-.r

( \ / AVERAGE WIND ROSES FOR OCTOBER '63-'64 w KEY WIND SPIEIED -1. 10 MPH -----11. 20 MPH 21.100 MPH w KEY WIND SPEED 1. 2 MPH 3 -6 MPH 7. 10 MPH N 2 o,,. E s TOTAL WIND N E s MUlllXVU...10 WIND < 10 MPH . Ul'IAR Aatif. 11 ftlev....., 186rn AVERAGE WIND ROSES FOR NOVEMBER 163-164 w KEY WIND SPEED -1. 10 MPH -----11. 20 MPH 21.100 MPH w KEY \'l'IND SPEED 1. 2 MPH 3. () MPH 7. 1 0 MPH N s TOTAL WIND N E E AVERAGE WIND ROSES FOR DECEMBER 163-164 KEY WIND SPEED -l. 10 MPH -----11. 20 MPH 21-100 MPH w KEY -WINO SPEED 1.002 MPH J.006 MPH 7. 10 MPH N s TOTAL WIND N E SSE s fl,Q\JRI xvu .. 12 WIND < 10 MP M UFIAll *.Rev. 11 (ttev.-. 1911'7) / w IC EY W:NO SPEED ------1-10 MPH 11-20 MPH 21-100 MPH ICE y '.VINO SP EE D 1-2 MPH 3 -6 MPH 7. I 0 MPH AVERAGE WI ND ROSES FOR '63-'64 20% E s TOTAL WIND N ENE s ,...._XW.1-3 WI.ND < 10 MPH ...... Aw. 1.5 tNov*mber 1917) +2 +1 -0 UJ 0::

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• L.LJ --a.. ('-! ::!E L.LJ I-<l 2 2 4 +2 +l --u.. 0 -..... L.LJ '"" a:::
• 0 I-I <( ...... 11111... a:::
• a.. ('-! LU I--1 -2 2 4 AVERAGE DIURNAL LAPSE RATE .... "'Ill "-. " --6 8 ' \. '\. ' ' ......_ 6 8 -10 12 14 HOURS MARCH '63-'64 ..1111 .... , _, ,, 10 12 14 HOURS APRIL '63-'64 "' .... ..11111 -"""'Ill 16 18 20 /f/lfll"' I ' , 16 18 20 ,.__xw .. 1.1 -..... 22 24 *--22 24 z 0 c;;; a:: UJ > z UJ en Q.. < ..J z 0 c;;; a:: UJ > z UJ en Q.. < ..J UFSM Aw. 11 Otevember 1997)

+2 .... -"""""" +l -l.J... 0 LU a::: 0 => ......, t-I ci: 0 a::: -a.. ('.,j :iE: LU t-<1 -I -2 2 -". +2 +l -l.J... 0 LU a::: 0 => ......, t-I 0 ci: a::: N LU= a.. ('.,j :iE: LU t-<1 -I -2 2 AVERAGE DIURNAL LAPSE RATE ......... ' ' l ' J 4 6 . B -..... '\ \ ' l , l \. ,. 4 6 B ....-' .... , / ' , , 10 12 14 HOURS MAY '63-'64 J Ill"" ' 10 12 14 HOURS JUNE '63-'64 ... ,,,,,.. ...._ J __ ,. ...... ..... , 16 18 20 22 24 /-J ,,_,, , " I I ...,j I 16 18 20 22 24 .,...xw .. 1a z: 0 c:;;; a:: LU > z: L&J V) Cl.. < _, z: 0 c;;; a:: L&J > z: L&J V) Cl.. < _, \INAR Rev. 15 i11evemher 1997) ' ' ./ +2 +l -LI.. 0 UJ a::: -I-I <( 0 a::: -1.1.J a... ('.J ::2:: 1.1.J I-<l 2 +2 +l -LI.. 0 1.1.J -a:::= :::::::> <"') I-I 0 <( -a::: ('.J 1.1.J = a... ('.J ::2:: 1.1.J I-<l 2 ...... "II "-. -2 4 .......... ---, 2 4 AVERAGE DIURNAL-LAPSE RATE ) I' ,, ,. -.. I , I I l , , ......... ' .... -.,. "' *-, ....... -----6 . 8 l \ \ , l \ \ ' ..... 6 8 10 12 14 HOURS JULY '63-'64 ..... 10 12 4 16 h 18 20 J , ' I .. / I 11 le HOURS PIGUftl XVU-17 22 22 24 ...... 2 4 z: 0 c:;;:; a::: ..... > z: z: 0 c:;;:; a::: Ll.J > z: ..... "" Q.. cc _, AUGUST '63-'64 UFIAR Rav. 11 thvtHf.lfMr 1M7) +2 +1 -LL.. 0 I.LI a::: -.0 :::> (¥') I-I <( 0 a::: N I.LI 0 a... C'..I ..., . ....:: I.LI I-<I 2 +2 +l -LL.. 0 I.LI a::: -I-I <( 0 a::: -I.LI a... C'..I :2= I.LI I- z: UJ V) a... <( ...J 6 8 10 12 14 16 18 20 22 24 \ 6 a HOURS SEPTEMBER '63*'64 ' ......... "' -,... 10 12 14 HOURS OCTOBER '63*'64 . , ...... J I I ' ..ii -, li 18 M 22 24 MUaXVl*11 z 0 c;;; a::: I.LI > z: I.LI V) a... <( ...J UPSAR Rev. 1& tllevamber 1997) AVERAGE DIURNAL LAPSE RATE +2 +l -u.. 0 ... ! ......... --1 -2 2 4 6 +2 +l -u.. 0 UJ 0:: 0 => CV> I-I < c....io 0:: UJ c:::> a... C'-.1 UJ I-<l ...... -._. "'"""""'" 2 2 4 6 December data based on '62 plus '63 data ('64 data unavailable because of instrument malfunction) ...._ . ..... ""' ...... '"' ..... ' 8 10 12 14 " ... " HOURS NOVEMBER '63*'64 \: 8 10 12 14 HOURS DECEMBER '62*'63 ..oil r' ... -" .... , .... ,,. 16 18 20 22 24 """""' _,_ .. .... " : ""' ., li 12 20 22 24 ..... JC'Vll.1.* z: 0 Vi 0:: uJ > z: uJ V> a... < ....J z: 0 Vi er:: uJ > z uJ V> a... < ....J *fftA11a.v.15 tNe:v.e.mur 1.9*97) 500 400 300 200 100 0 ) LAPSE RATES BY WI ND SPEED AND TURBULENCE CLASSES FOR JANUARY 163-164 ... .. .. --.. ---------------I ii ::: ... -... i-: : :-: ---TEMPERATURE DIFFERENCE 202' -30' +5.0 TO +35 +0.6 TO +4.9 -0.5 TO +0.5 0 l.O TO -0.6 0l5TO-l.l KEY SYMBOL ---* * *=* 203' WIND SPEED GROUP A 0 TO 10 MPH GROUP I GROUP C II TO 20 MPH 21 TO 100 MPH :*-: . . *=* *=* :: :*: . :*: * * *=* ::: --: ::: ::: ::: ________________ : =.--: :. : __ ;: :-. *.* .*. --* * * *

• A B CLASS c . . . . . . . : . . . .. *. . . . . . . . . .... . . . . . ... : . : . . ... * . . . . . . . . * ::: A . . . . . *. *. *. *. *. * . . : . . : . ::: ::: ::: :*: . . . . -. . . *=* .*. .. . . . *=* .*: "!-"'\ ::: ::: :=: :*: .*.* .*. *.* *=* ::: . . .. . . . . . . . B CLASS II c A B C CLASS Ill AGUMXV.-* --

--*

• A --B CLASS IV --c UFSM Rev. ti fhwmber 1997)

( 400 300 11'1 1:111: ::::> 0 % 200 100 . . . *. 0 .

• A LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR FEBRUARY '63-'64 KEY TEMPERATURE DIFFERENCE SYMBOL 202* -30' +5.0 TO +35 +0.6 TO +4.9 -0.5 TO +0.5 -1.0 TO
• 0.6 -* .15 TO
• t.I * *=* 203' WIND SPEED OllOUP A 0 TO 10 MPH GllOUP I I I TO 20 MPH * * * -* -. * -* * * -OllOUI' C 21 TO 100 Ml'H * * * * * * * * * . . * * . * . . . * -* * * * . . * --. . ! . . * * . . * * * . . * * . . . . **--* . . --. . . * * ** * * * * * * * * * * * ** * * * * * * . * * * ** * * * * * * * * * :*: ** * * * * * ** . * * . * * * * . . --= ...... . . --. -. . . . . B c A B c A B c A B CLASS I CLASS II CLASS Ill CLASS IV PtGUlll XV&l--11 UFIAll llav. 11th,,.,... 181'1) c

.J "' -;:) 0 :c 500 400 300 200 100 0 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR MARCH 163-'64 ----------TEM,EIATURE DIFFERENCE 202*-30' +5.0 TO +35 *0.5 TO +0.5 -1.0 TO -0.6 *15 TO *1.1 KEY SYMBOl ---*** *** 203' WIND 5'EED OllOU' A 0 TO 10 M'H OllOU' I 11 TO 20 M'H OllOU' C 21 TO 100 M'H i ll! I ... .*. ... *=* *=* . . . . . . . . . '" :*: :*: :-: ... '----------------.: .--.: .-. : .--------:. :: . .. . .. . . '" ... ... *:* . . . . . . . . . . . . . . JgQC . . A B C CLASS I * * * * * * * * * * * * :*: :*: :-: ...... *.* *=* *=* *=* *=* = ::: .:.::. :*: :*: :*: = *:* . . . . . . . ::: ::: ::: *=* *=* ::: A C A B C CLASS II CLASS Ill PIOUMXW-D ------! -----*** ** *** * * . A B CLASS IV UF8M .... ti Uhw ..... ,,.,. .. . . c ( / "' DC ::> 0 ::z: 500 400 300 200 100 0 -* * * *** A LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR APRIL 163-164 -._.. . . . * * . *=* . . *** *** * . . * *

• B CLASS I c . . ** A a.JI -------B CLASS II c * * * * * * * * * * * * * * *
• A B CLASS Ill KEY TEMPHATUIE DIFFHENCE SYMIOL 202'-30' +s.o TO +u +0.6 TO +4.t ----0.S TO +o.s * -1.0 TO -0.6 -IS TO -U *** *** 203' WIND SPIED GROUP A 0 TO 10 MPH GROUP I 11 TO 20 MPH GROUP C . . . . . * * * * * . . ... c 21 TO 100 MPH --------. . . . A 8 CLASS IV FIGURE XW..23 UFSU ""* 11 {tf9Yen-.r tl9J) c LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR MAY 163-164 ' ' 500 KIT TlMPHATUH OIFFEHNCE SYMIOl 202*-30' +s.o To +u -+0.6 TO +4.9 ---0.S TO +o.s
• 400 -1.0 TO -0.6 *IS TO *1.1 * * ***
• 203' WIND SPIED GROUP A 0 TO 10 MPH GIOUP I 11 TO 20 MPH GROUP C 21 TO 100 MPH 300 .-------"' -llllC -::> --0 ::c ----200 ----§ ----

- --*----I *----100 ll ! * . -. * . * . * * -. * . . * * * * . --. . * * * * . ---. . -* ** . . . -* . . * ** . . m . . . -. * . . *** * . . . . . . . . *** ... . . . . * ** . * * . . . . * . . . . . . * . . . . 0 A 8 c A 8 c A 8 c A 8 c CLASS I CLASS II CLASS Ill CLASS IV / "8UMXW-2A UPIAR Rev. 11 fNev_.., 167) ( 500 400 300 200 100 . . . . . . 0 A LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JUNE '63-'64 ----------. * . * * . . * . . . * . . . . . . . . . . . . . ,.,.., . * . . . . B c A B c CLASS I CLASS II --* * * * * * * * * '797

• A B CLASS Ill ICEY TEMPEIATUH DlfFEIENCE SYMIOl 202*-30' +s.o TO +H +0.6 TO +4.t ----o.s TO +o.s * -1.0 TO -0.6
• 15 TO -t.I ** ***
• 203' WIND SPEED OIOUP A 0 TO 10 MPH OIOUI' I OIOUI' C -* . * . . :*: . . * . * * * * * * * * ***
• c II TO 20 MPH II TO 100 MPH s ----------------------i . . . A B CLASS IV Ml*XW-21 c UFIAll .... ti Woveanltaf 1811

( 500 400 300 200 100 0 / ---------.. .. -.. .. .. .. .. -... .. ---... .. --... LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR JULY 163-164 --------- -

---

* I : * --------TEMPERATURE DlfFllENCI 201*-:SO' +s.o TO +u +0.6 TO +4.9 -o.s TO +o.s

• 1.0 TO 00.6 *UT0-1.1 KEY 203' WIND SPEED OllOUP A 0 TO SYMIOL ---*** *** 10 MPH ----GROUP I GROUP C 11 TO 20 MPH 21 TO 100 MPH ------------I = ::: :*: = = ___________________________ .... =*= m = --------. . . . . . . A . . * * . . . B CLASS I . . c .*. *=* ---= ::: ::: -= : : : : : : : : : . = .*. *:* . **: . . . = .*. . . :*: :* *.* -. . . **:* .* . . . -. . . . . . : . . . . -: . : : : : . . : . : . : . I . . . . . . . . . . . . . . . : . : . : . : . ..... . . : . : : : '* . . . .. . .. . . . . -. . . . . . . ... A B CLASS II c A B CLASS Ill c AGUMXVU.28 A B CLASS IV c *IJFUA ltav. 11 llk>wmbar 19171

( "' :::> 0 :c soo 400 300 200 100 0 --------.. -.. -.. .. .. -.. --..... . . . . . . * . . . . . * * * . . A LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR AUGUST '63-164 -. -B CLASS I c --= A -----

---

--- -- --I B CLASS II c A B CLASS Ill TEMPERATURE DIFFERENCE 202' -30' +5.0 TO +35 +0.6 TO +4.9 -0.5 TO +0.5 -1.0 TO -0.6

• 15 TO
• f.1 KEY SYMBOL ---*** *** 203' WIND SPIED GllOUP A 0 TO 10 MPH GllOUP I GllOUP C c 11 TO 20 MPH 21 TO 100 MPH A -

---B CLASS IV ..... XW-27 c WU.-.. tS Clttt_.* 1Ull $00 *oo 300 200 100 0 ... ... ... ... LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR SEPTEMBER '63-'64 ------ -

---

-TEMPERATURE DIFFERENCE 202'-30' +s.o TO +u +0.6 TO +*.9 *0.$ TO +0.5 -1.0 TO -0.6 *15 TO* I.I ICEY SYMBOL -----203' WIND SPEED -GltOUP A 0 TO 10 MPH --... ----------GROUP I GIOUP C II TO 20 MPH 21 TO 100 MPH -I -: -i .. .. ----.... : .... ti-* * * -. . . -a-=*.: *** = ...

• c *** -*:* *=* = = . . *.* --... *:. . :* *.* .... . . . . . . '"'\,: Ii-.... * * *** -. -. . . . . . . .. . *=* :*: :*: : . : . . . . .. . . . . : . . . . . .. :*: .*. . .. ::: :* .* : ::: ::: * ;,..llL * : * * * :*.:_ ..... :*: *.* *=* ... . -. . . . . . . .. --A B C A B C CLASS I CLASS II --: ::: . . * * ** A B C CLASS Ill FIGURE XVll-28 --------------..,. A B C CLASS IV UFSAR Rev. 15 (November 1997)

( \ 500 400 300 200 100 0 LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR OCTOBER '63-' 64 ----------------------:....*515 --------------------

---

I .. = . : TIM,EIATUIE DlfFllllNCE 202'-30' +5.0 TO +3S +0.6 TO +4.t -o.s TO +o.s

• 1.0 TO *0.6 *15 TO *I.I ICEY SYMBOL ---** *:* 203' WIND .,HD OIOU' A 0 TO 10 M'H OIOU' I OIOU' C II TO 20 M'H 21 TO 100 M'H : --------------! I ! rn ... -. . . -*.* ... -... . . = *.* . . -*.* . . = .... -*.* . . = .. . *.* . . --. *. *.* . . I . . . . . . . . -. . . . . . . . . -*=* :*: .. . . . . . . . .. ... . . . .. . . . . -= :*: *=* :*: .*. . . . .*. . .. ':*. *=* :*: *=* --.: ... .. . . .. . . . . . -.=.*.= ... . .. : . . . . . ... :-'\ ... .*. *.* .*. ...* .. *.* ----

---

-... A B C A B C A B C A B C CLASS I CLASS II CLASS Ill CLASS IV AGUlll XVll*D UFIAll Rav. 11 t.897) ( soo 400 300 200 100 0 / LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR NOVEMBER '63-'64 '508 ----TEMPERATURE DIFFERENCE 202' -30' +5.0 TO +35 --+0.6 TO +4.9 ----00.5 TO +0.5 -1.0 TO 00.6 0l5T0°1.1 KEY SYMIOL --* * * *=* 203' WIND SPEED ----------GllOUP A 0 TO 10 MPH GllOUP I GllOUP C 11 TO 20 MPH 21 TO 100 MPH ::: *** *

• A B CLASS I c *** . . *** * . . *** . *** * * *** *** A '""' ... ::: :*: .*. .*. ... *:* * * * . . .. . . . --. . . . . . * * ** . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . * *
• B CLASS II . . * . . * . . * . . . . . . . * . . . . .
• c }* . * * ** --* * * * -:.:: * * ** * ** -A B c A B c CLASS Ill CLASS IV "8UlllMl-IO UPSAI st:ev. 11(tloveBther111lt

( **--.,,. .soo 400 300 200 100 0 --------------------... ------LAPSE RATES BY WIND SPEED AND TURBULENCE CLASSES FOR DECEMBER '63-'64 = :*: *:* TEMPEIATUIE DIFFERENCE 202'-30' +s.o To +35 +0.6 TO +4.9 .o.s TO +o.s 0 l.O TO *0.6 01.STO*l.1 KEY SYMBOL ---* * *=* 20:1' WIND SPEED GROUP A 0 TO IO MPH GROUP I GROUP C II TO 20 MPH 21 TO 100 MPH -.*. *.* 1----------------0---*: *--.. *---------------------------.... -... ... -... .. ::: :*: :*: :*: :*: ::: ::: ::: 1----------------. :--.. ._ . --------* .,------------------1 *.* *.* ... *.:*. .. -------. . * *** * * * * * * *

• A B C CLASS I *.* *.* ... . *.* *.* :-: *.* :*: ::: *=* :-: ... . . . . .* . . *. *=* ::: *=* . : . . . . . .. .. . * .. ** . . . . .. -.. . . . . . .*. -.. . :*: ::: *=* = :*: .::.: ::: :*: ::: . . . . *. :. *. . ... . . . .. . . . . . . -. . . . . . . . A 8 C A B C CLASS II CLASS Ill ----.. ---........, A 8 C CLASS IV December data based on '62 plus '63 data ('64 data unavailable because of instrument malfunction) FIGURE xvu .* 31 UFIAflRav.

( Lake Onta:rio . _.,,I o, UTICA -.... \ I 'I \ I "'t-( 10. 8 0.1 j I I I I I I CENTERLINE CONCENTRATIONS II. I I I I I I I ' I I I I I I I I I\ II. TURBULENCE CLASS I TRAJECTORY FROM WATER \ WIND: 2 M/SEC \ LID: 1500 M RELEASE: 1 UNIT/SEC -"""" 'ii-. II. ' \. '\. \. '\ \ ' ' \ \ \ ' \GROUND SOURCE ELEVATED ' \ ' ' ' " \. " "" I\. \. ' " " "' "' " """ 1.0 10 100 DISTANCE (KM) AeURIXW .. U UNAa ..... 11 , * .,,. ( CENTER LI NE CONCENTRATIONS I I I I I I ' I I I I I ' ' TURBULENCE CLASS II \ TRAJECTORY FROM WATER \ WIND: 4 M/SEC \ LID: 1500 M RELEASE: 1 UNIT/SEC --,... ""'r-. \ J \ I ' ' , ' '\. " ' J '\. ' I ' " I I\ \ I \ ' I , \GROUND SOURCE \ ELEVATED ' \ " " ' '\. , \ ' ' " ' \ ' \ ' ' \ \. " 8 10. 0.1 1.0 10 100 DISTANCE (KM) FIGURE XVU-34 UF&AR Rev. 11(NoY&mb.er1997) CENTER LI NE CONCENTRATIONS " ' \ ' TURBULENCE CLASS Ill ' TRAJECTORY FROM WATER ' WIND: 6 M/SEC ' LID: 500 M ' RELEASE: 1 UNIT/SEC \ ' ' " " ' \ \ ' GROUND SOURCE \ \ io* 6 \ ' 7 ""'-" / "' 7 !'-. ' 7 ' ' ' I ELEVATED SOURCE\ ' \ 7 1.0 10 100 DISTANCE (KM) FIGURE UfUll Rev. 19871 ( CENTERLINE CONCENTRATIONS ' ' \. " ' TURBULENCE CLASS IV " "'-TRAJECTORY FROM WATER \ WIND: 2 M/SEC \ LID: 200 M RELEASE: I UNIT/SEC ' ' " \. ' \ ' GROUND SOURCE \ ' j\ 10. 5 ' ' ' "' \. \ \ (ELEVATED SOURCE < 10-6) \ 6 \ 1.0 10 100 DISTANCE (KM) AGURI xv111 .. 3e UflAR Rev. 11 Pll*vallllber 1a1Jfl) / 10 -8 0.1 I I I I I I I , CENTERLINE CONCENTRATIONS I I I I I ' . TURBULENCE CLASS Ii _._ .... ............ \ BECOMING --....... \ CLASS IV --...... \GROUND SOURCE AT 2 KM AND ----\ CLASS II AT 23 KM ----WIND: 3M/SEC -..... --i,.. """' r---. LID: 1500 M t--. ...... RELEASE: lUNIT/SEC (SECTOR CONLY) """""" ' ..__ ' --. ' ' ' ' --.. ............._ l """" ELEVATED SOURCE' 1.0 10 100 DISTANCE (1<M) NGUREXW-37 UNM Rev. Hi t"'Gvamtler 1881) ( ' / 10*7 0.1 CENTERLINE CONCENTRATIONS I I I I I I I I'\. I I I I I II '\. I '\. I I I I I I I I '\. TURBULENCE CLASS IV BECOMING I\. CLASS II ' AT 16 KM \ WIND: 2 M/SEC LID: 1500 M ' RELEASE: 1 UNIT/SEC \ (APPLIES TO SECTOR D1 GROUND SOURCE I\ DAYTIME) , \. \. \ l ELEVATED SOURCE \ 1.0 10 100 DISTANCE (KM} NUMXVl-31 LWIAll ftw. 11 (**emtter 1;M1t ( \ / 10* 7 0.1 CENTERLINE CONCENTRATIONS I I I I I I I TURBULENCE CLASS IV BECOMING CLASS II AT 2 KM WIND: 2 M/SEC i LID: 1500 M -RELEASE: !UNIT/SEC (APPLIES TO SECTORS D2 TH RU G DAYTIME): _.,.. ... ---l ' ' I \ \GROUND SOURCE ' \ ' \ .. \ \. ' \ ' ' ' \. ELEVATED SOURCE ,, ' \. Ii. l.O 10 100 DISTANCE (KM) "8URI XVA*31 UF8M ... 11 tNewttlhf t991I ( 10" 6 10-7 9 10" 0.1 ' I J ' ' ' I """' ' I !/ r RADIAL CONCENTRATIONS ' ' ' TURBULENCE CLASS I \ TRAJECTORY FROM WATE.R WIND: 2 M/SEC ,\ LID: 1500 M RELEASE: 1 UNIT/SEC I\. ' ' ' " , ' ... \ \ \ \ \ ' GROUND SOURCE \ \ \ , ELEVATED SOURCE \ \ 1.0 ' ' ' I'-' \ , ' 'I' ' 'I ' ' "' ' 10 100 DISTANCE (KM) RGUMXW-40 ( ' / 10* 7 10" 8 9 10" 0.1 J I I ' ' , ' ' I ' I I RADIAL CONCENTRATIONS I I I I I I I I I I I ' ' TURBULENCE CLASS II .......... --TRAJECTORY FROM WATER ........ i\ WIND: 4 M/SEC .......... LID: 1500 M ' --RELEASE: !UNIT/SEC \ \ ' ,, ' '-" , '\. \ \. \ \ \ ' , \GROUND SOURCE \ ELEVATED \ \ \ ' ' ' ' ' " 1 \. \ 1 \ \ 1 \ \ ' ' ,\ ' 1.0 10 -100 DISTANCE (KM) RQ:LJM XV;l4 'I W&All 118¥. 11(November1887) ( RADIAL CONCENTRATIONS ' I\. ' " TURBULENCE CLASS Ill ,_ .... TRAJECTORY FROM WATER >-'-\ WIND: 6 M/SEC >-'-\ LID: 500 M RELEASE: l UNIT/SEC --\ io* 7 ' ' \. \ ' \ \ -SOURCE / ' ' i\ 8 \ ' " io* -i \ \ , ' \ ' " I ELEVATED SOURCE' \ ' \\ .9 10 0.1 1.0 10 100 DISTANCE (KM) flGUBI XVU-42 WIAI Rev. 11 1997) c* , 10 -8 0.1 RADIAL CONCENTRATIONS I I I I I ll I I I I I " I I I I I TURBULENCE CLASS IV ........

• 1-1-"' TRAJECTORY FROM WATER \ ._._ WIND: 2 M/SEC \ LID: --'-200 M RELEASE: 1 UNIT/SEC \ \ ' ' ' '\ ' \ \ GROUND SOURCE \ ' I'-' ' \. (ELEVATED SOURCE < 10-8) \ ' \ \ \ 1.0 10 100 DISTANCE (KM) PlGIJlll XW.._3 YPIAB hv. 11 !NG¥Ullti* 11971

/ \ RADIAL CONCENTRATIONS I I I I ' I I I I ' I I I \ TURBULENCE CLASS II ' ' BECOMING CLASS IV AT 2 KM AND \ CLASS II AT 23 KM \ WIND: 3M/SEC LID: 1500 M RELEASE: 1 UNIT /SEC 7 ' (SECTOR CONLY) 10* .. " '" , . " " ' "' " I '\ " , ' \ ' i\. f' SOURCE I ELEVATED SOURC} " r\. 8 ... .... 10* I' " "' '" " " "" "' ' \ \ \' ' \ 9 ' 10*0.1 1.0 10 100 DISTANCE (KM) AGUBEXW......, WaM Ile¥. t:t (ftovember 18:811 ( \ RADIAL CONCENTRATIONS "' I I I T T , I I I T I l r I I 1 ' , TURBULENCE CLASS IV -\ BECOMING -'--'-'"-\ CLASS 11 -'--'-'"-AT 16 KM --1-1-\ WIND: 2 M/SEC -'-........... \ LID: 1500 M RELEASE: 1 UNIT SEC \ (APPLIES TO SECTOR 01 DAYTIME) -..... _ ..... ' " ' , \ \ ""' l ' GROUND SOURCE ' 7 " ' lo* ' \ \ \ \ (ELEVATED SOURCE< 10-8) 8 ' 1.0 10 100 DISTANCE (KM) ftGUtlE XVB-41 WIM Rev. 11 {Hove*-11117) ( RADIAL CONCENTRATIONS " ' \ TURBULENCE CLASS IV I\ BECOMING \ CLASS II \ AT2 KM \ WIND: 2 M/SEC LID: 1500 M RELEASE: 1 UNIT/SEC (APPLIES TO SECTORS D2 THRU G DAYTIME) . ' \GROUND SOURCE ' 7 ' 10* ' \ \ ' \ ' l \ ' ELEVATE! SOUiCE ' 8 \ \. 10"0.1 1.0 10 100 DISTANCE (KM) A8UMXVl*46 UN.Ul .., . 11 (,Neva..., 1197) LLJ <( 0:: LLJ en 0 Cl 10 0. 1 1

• CENTERLINE GAMMA DOSE RATES I I I I I I I I 1 ' I ""-.. TURBULENCE CLASS I --..... '-" .. " TRAJECTORY FROM WATER --..... '---" WIND: 3.5 M/SEC LID: 1500 M -'--'-'--\ RELEASE: l Ci/SEC \. \ I I ... ' ' .., \ ELEVATED SOURCE ' 'I " ' " \. ' \. " 1.0 10 100 DISTANCE (KM) FIOUJll XVl,-4 7 *taau ....,. 11 t987)

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NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Section XVIII XVIII-i Rev. 25, October 2017 SECTION XVIII HUMAN FACTORS ENGINEERING/SAFETY PARAMETER DISPLAY SYSTEM A. DETAILED CONTROL ROOM DESIGN REVIEW 1.0 General 2.0 Planning Requirements for the DCRDR 3.0 DCRDR Review Process 3.1 Operator Survey 3.2 Historical Review 3.3 Task Analysis 3.4 Control Room Inventory 3.5 Control Room Survey 3.6 Verification of Task Performance Capabilities 3.7 Validation of Control Room Functions 3.8 Compilation of Discrepancy Findings 4.0 Assessment and Implementation 4.1 Assessment 4.2 Implementation 4.2.1 Integrated Cosmetic Package 4.2.2 Functional Fixes 5.0 Reporting 6.0 Continuing Human Factors Program 6.1 Fix Verifications 6.2 Multidisciplinary Review Team Assessments 6.3 Human Factors Manual for Future Design Change 6.4 Outstanding Human Factors Items 7.0 References B. SAFETY PARAMETER DISPLAY SYSTEM 1.0 Introduction to the Safety Parameter Display System 2.0 System Description 3.0 Role of the SPDS 4.0 Human Factors Engineering Guidelines 5.0 Human Factors Engineering Principles Applied to the SPDS Design 5.1 NUREG-0737, Supplement 1, Section 4.1.a NMP Unit 1 UFSAR Section Title Section XVIII XVIII-ii Rev. 25, October 2017 5.1.1 Concise Display 5.1.2 Critical Plant Variables 5.1.3 Rapid and Reliable Determination of Safety Status 5.1.4 Aid to Control Room Personnel 5.2 NUREG-0737, Supplement 1, Section 4.1.b 5.2.1 Convenient Location 5.2.2 Continuous Display 5.3 NUREG-0737, Supplement 1, Section 4.1.c 5.3.1 Procedures and Training 5.3.2 Isolation of SPDS from Safety-Related Systems 5.4 NUREG-0737, Supplement 1, Section 4.1.e 5.4.1 Incorporation of Accepted Human Factors Engineering Principles 5.4.2 Information Can Be Readily Perceived and Comprehended 5.5 NUREG-0737, Supplement 1, Section 4.1.f, Sufficient Information 6.0 Procedures 6.1 Operating Procedures 6.2 Surveillance Procedures 7.0 References NMP Unit 1 UFSAR LIST OF TABLES Table Number Title Section XVIII XVIII-iii Rev. 25, October 2017 XVIII-1 SPDS PARAMETER SET NMP Unit 1 UFSAR Section XVIII XVIII-1 Rev. 25, October 2017 SECTION XVIII HUMAN FACTORS ENGINEERING/SAFETY PARAMETER DISPLAY SYSTEM A. DETAILED CONTROL ROOM DESIGN REVIEW 1.0 General Nine Mile Point Nuclear Station - Unit 1 (Unit 1) initiated a control room review program in response to NUREG-0660, "NRC Action Plan Developed as a Result of the TMI Accident"; NUREG-0737, "Clarification of TMI Action Plan Requirements"; and NUREG-0737, Supplement 1. In 1982 the Nuclear Regulatory Commission (NRC) issued orders requiring licensees to perform a detailed control room design review (DCRDR) to identify and correct design deficiencies. NUREG-0700, "Guidelines for Control Room Design Review," issued in September 1981, provides human engineering guidelines to assist each licensee and applicant in performing a DCRDR. Unit 1 carried out a DCRDR in two major phases. The initial phase was an early human factors review under the auspices of the Boiling Water Reactor Owners' Group (BWROG) in accordance with the methodology they developed for conducting control room design reviews, as described in their program documentation, NEDC-30285, "BWR Owners Group Control Room Design Review Program." The initial review was completed July 1981, the resulting human engineering observations (HEOs) were compiled, and a general assessment was carried out which verified that there were no severe problems requiring immediate resolution. The second phase of the program was initiated in response to the follow-on NRC guidance provided in NUREG-0737, Supplement 1. The HEOs identified in the first phase of the DCRDR were included for reassessment and resolution in the second phase. ARD Corporation was contracted as human factors consultants to perform the second phase of the DCRDR. The DCRDR process, as suggested by NUREG-0700, was divided into four major activities: planning, review, assessment and implementation, and reporting. In addition to the completion of these initial DCRDR activities, Unit 1 has made the DCRDR a continuing activity to ensure that the human factors improvements made by the DCRDR program are carried on throughout future control room design modifications. The human engineering NMP Unit 1 UFSAR Section XVIII XVIII-2 Rev. 25, October 2017 processes developed to address the Unit 1 DCRDR requirements are described in Sections 1.1 through 1.4. 2.0 Planning Requirements for the DCRDR The Unit 1 Summary Report and Program Plan was submitted on September 30, 1983. This document included the results of the initial phase of the review and the outline for the final phase to complete the DCRDR program. The program plan described the of the DCRDR, the composition of the multidisciplinary review team to carry out the review, data management processes to be used, and the schedule of activities to be performed. 3.0 DCRDR Review Process The review process focused on whether the control room provides the information and control capabilities necessary for control room Operators to accomplish their functions and tasks effectively. The process identified characteristics of the existing control room instrumentation, controls, and other equipment and physical arrangements that may detract from Operator performance (Final Summary Report for Detailed Control Room Design Review of Nine Mile Point Unit One, July 1985, Section 4.0). 3.1 Operator Survey The purpose of the Operator survey (Final Summary Report Section 4.1) was to bring out the unique knowledge of control room operating personnel about the problems and positive system features that they have noted in their operating experience. A questionnaire covering each of the nine topic areas of NUREG-0700 was distributed to all Unit 1 licensed Operators. Ambiguities in the written responses to the self-administered questionnaire were further addressed in follow-up interviews conducted by human factors personnel on site. Nineteen Operators responded to the questionnaire. The responses revealed, in general, favorable attitudes about the control room. The responses to the questions did not reveal a majority of strong negative concerns on any single question. Some concerns were defined well enough to be written as HEOs. 3.2 Historical Review The objective of the historical review (Final Summary Report Section 4.2) was to identify problems previously encountered NMP Unit 1 UFSAR Section XVIII XVIII-3 Rev. 25, October 2017 involving control room operations, and to investigate the potential enhancements to the control room which could significantly reduce the potential for human error. The documents investigated were: Licensee Event Reports (LERs) for Unit 1, scram reports for Unit 1, and selected LERs, Significant Operating Event Reports (SOERs), and Significant Event Reports (SERs) found to be significant to BWR plants. Reports were screened to determine if they described and documented a control room problem related to human factors concerns. A Problem Analysis Report (PAR) was prepared for every report that cleared the initial screening. Two HEOs were identified for DCRDR consideration. 3.3 Task Analysis The task analysis (Final Summary Report Section 4.3) was performed to determine if the control room instrumentation and controls are adequate for the tasks Operators are required to perform in emergency situations. The Plant-Specific Technical Guidelines (PSTG) and Plant-Specific Severe Accident Guidelines (PSSAG) were used to derive detailed information on Operator response during transients and accidents. All steps and contingencies represented in the PSTGs/PSSAGs were analyzed, resulting in a list of plant-specific tasks for the accomplishment of all branches of the guidelines. Each task was generally comprised of several subtasks or action steps. Detailed data regarding the information, control, and feedback characteristics of the instrumentation required to perform each action step within the task was recorded. The task analysis data constitutes a specification of Operator needs to accomplish the Operator functions. This specification was used as a foundation reference point to verify the availability and suitability of control room instrumentation. 3.4 Control Room Inventory The inventory (Final Summary Report Section 4.4) identified all control room instrumentation, controls, and equipment which served as a reference for comparison with the requirements identified in the task analysis. The inventory data was collected on a panel-by-panel basis. All information relevant display range and units, control switch positions, and annunciator legends, was recorded and stored in a computerized database. 3.5 Control Room Survey NMP Unit 1 UFSAR Section XVIII XVIII-4 Rev. 25, October 2017 The control room survey (Final Summary Report Section 4.5) is a methodical comparison of control room design features with established human engineering guidelines. The Unit 1 survey was performed in two stages. The original survey was conducted in July 1981 using the checklist in the BWROG Control Room Survey Supplement. In response to an in-progress audit by the NRC in late 1984, an additional survey was performed using the NUREG-0700 Section 6 guidelines. 3.6 Verification of Task Performance Capabilities The objective of the task performance verification (Final Summary Report Section 4.6) was to ensure that Operator tasks could be performed in the existing control room. The verification compared the information and control needs, and their required characteristics identified in the task analysis effort, to the corresponding control room inventory components and characteristics to verify the availability and suitability of control room equipment to support Operator task performance. The verification process was a computerized matching of corresponding data fields. Manual verifications were performed for back panel and other instrumentation whose characteristics were not fully described in the inventory database. A HEO was generated if the task analysis stated a need for a component that was either unavailable or unsuitable. 3.7 Validation of Control Room Functions The validation (Final Summary Report Section 4.7) was performed to determine whether the functions allocated to the control room operating crew could be performed using the EOPs within the design of the control room. Two validation methods were used, a walk-through and a talk-through. For the talk-through method, four different accident scenarios were carried out in the Unit 1 simulator. The scenarios involved the operating crew performing the activities necessary to mitigate the accident. Their actions were videotaped with a voice-over narration explaining their actions. Human factors specialists analyzed the videotapes to determine that each step of the procedures could be performed within the boundaries of the control room design criteria. The talk-through method involved an Operator demonstrating to a human factors specialist, in the Unit 1 simulator, the tasks and equipment responses for each of the four selected events and all tasks identified in the Unit 1 task analysis effort. Specific plant equipment and Operator NMP Unit 1 UFSAR Section XVIII XVIII-5 Rev. 25, October 2017 decisions involved in each task were identified as the Operator described the actions from the applicable simulator workstation. 3.8 Compilation of Discrepancy Findings Each of the review processes described identified HEOs. The HEOs were compiled in a database and identified according to the review process from which they were identified (Operator survey, checklist survey, etc.). Each HEO had a unique identifying number and a description of the deviation from the human factors guidelines. 4.0 Assessment and Implementation The DCRDR assessment refers to the process of evaluating the HEOs identified during the review process. The implementation describes the activities that were performed to correct the HEOs that were determined to be possible sources of human error. 4.1 Assessment During the review process of the DCRDR, 530 HEOs were identified. The objective of the assessment (Final Summary Report Section 5.2) was to analyze and evaluate the problems that could arise from the HEOs. The multidisciplinary team that was responsible for performing the DCRDR review served as the assessment team. Each HEO was evaluated separately. They were categorized by functional category, safety significance, and risk potential. Finally a decision was made, based on these factors, regarding the resolution; whether the HEO should be fixed, not fixed, or was invalid (a misunderstanding of the guidelines or control room conditions). 4.2 Implementation The assessment established modifications and a schedule to address the HEOs that were resolved to be fixed. The implementation of the fixes to correct the problems identified by the HEOs was divided into two basic packages, the Integrated Cosmetic Package (ICP) and functional fixes. 4.2.1 Integrated Cosmetic Package The ICP (Final Summary Report Section 5.3.2) was a series of changes to enhance the appearance and the ability of Operators to effectively locate components on the Unit 1 control panels. Cosmetic changes are those which alter operational aids; these NMP Unit 1 UFSAR Section XVIII XVIII-6 Rev. 25, October 2017 included enhancements to the overall appearance of the control panels to assist the Operator in determining component location and status. The ICP did not modify system operation, only the information on the control panels presented to the Operators to assist in performance of their assigned duties. The ICP consisted of the following enhancements: 1. System demarcation lines and system/subsystem labeling. 2. Component labeling. 3. Indicator scale replacement. 4. Indicator scale range marking. 5. System mimics. 6. Control handle conventions. 7. Indicator light color. 8. Recorder chart paper. Questionnaires were distributed to two operating crews in the simulator to obtain feedback on Operator response to the ICP. The results of the verification were inconclusive. The result was that the assessment team came to a consensus agreement about the type and scope of each fix to be applied. The execution of the details (size, placement, etc.) was completed as the engineering packages were assembled. A verification investigation of the ICP was performed in the Unit 1 simulator. The changes specified in the ICP were installed on the simulator control panels to verify that the proposed cosmetic changes were comprehensive, accurate, and that the changes did not create any new problems. 4.2.2 Functional Fixes Functional fixes were those modifications, performed to correct DCRDR HEOs, that affect the functional operation of the control room, such as training, cracked control handles, or inoperative annunciators. A verification of all functional fixes was performed by human factors specialists to verify that the installed modifications effectively corrected the identified HEOs and did not cause any new problems. NMP Unit 1 UFSAR Section XVIII XVIII-7 Rev. 25, October 2017 5.0 Reporting Unit 1 has maintained regular communication with the NRC regarding the progress of the DCRDR. The following is a summary of the DCRDR reporting: September 1983 Unit 1 submitted a Summary Report and Program Plan February 1984 NRC provided Unit 1 with comments on the DCRDR Program Plan November 1984 The NRC conducted an in-progress audit of the DCRDR January 1985 The NRC provided Unit 1 with a report on the in-progress audit May 1985 Unit 1 personnel met with the NRC staff to discuss concerns identified during the in-progress audit July 1985 Unit 1 issued a Final DCRDR Summary Report July 1986 NRC issued a Safety Evaluation Report based on the Final Summary Report January 1987 Unit 1 issued a Supplemental DCRDR Summary Report June 1989 The NRC conducted an onsite DCRDR audit August 1990 NRC issued a Safety Evaluation Report based on the DCRDR 6.0 Continuing Human Factors Program When the majority of the work on the DCRDR was complete, Unit 1 established a formal human factors program. The human factors program has verified the completion of human factors related control room modifications, held periodic assessments of potential human factors problems, created and updated a Human Factors Manual, and updated the task analysis to incorporate revisions to the EOP flowcharts. 6.1 Fix Verifications NMP Unit 1 UFSAR Section XVIII XVIII-8 Rev. 25, October 2017 Modifications related to human factors problems, either identified as HEOs or through other review processes, are verified as complete by human factors specialists. The verification includes ensuring that the modification sufficiently addresses the identified problem and that it does not introduce any additional problems. 6.2 Multidisciplinary Review Team Assessments Unit 1 has periodically convened a multidisciplinary review team to address outstanding human factors problems or issues. The review team is comprised of representatives from operations, engineering, training, and human factors specialists. 6.3 Human Factors Manual for Future Design Change A Human Factors Manual for Future Design Change has been prepared for Unit 1; Revision 4 was distributed in June 1991. The human factors manual establishes plant-specific human factors conventions based upon NUREG-0700 guidelines. 6.4 Outstanding Human Factors Items There are several outstanding HEOs scheduled for implementation during the 1992 and 1994 outages. All of the outstanding HEOs are functional fix items that have involved extensive engineering work. 7.0 References 1. NUREG-0660, NRC Action Plan Developed as a Result of the TMI Accident. 2. NUREG-0737, Clarification of TMI Action Plan Requirements. 3. NUREG-0737, Supplement 1, Requirements for Emergency Response Capability. 4. NUREG-0700, Guidelines for Control Room Design Review. 5. NEDC-30285, BWR Owners Group Control Room Design Review Program. 6. NMPC letter to NRC, April 15, 1983, Response to NRC Request for Information Regarding Supplement 1 to NUREG-0737 (Generic Letter 82-33). NMP Unit 1 UFSAR Section XVIII XVIII-9 Rev. 25, October 2017 7. NMPC Summary Report and Program Plan, September 30, 1983. 8. NRC letter to NMPC, February 22, 1984, Staff Comments on the Detailed Control Room Design Review Program Plan. 9. NRC letter to NMPC, October 1, 1984, Meeting Summary - Detailed Control Room Design Review - August 17, 1984. 10. NRC letter to NMPC, January 1985, Report on the Detailed Control Room Design Review In-Progress Audit. 11. NRC letter to NMPC, July 3, 1985, Meeting Summary - Detailed Control Room Design Review (DCRDR) - May 9-10, 1985. 12. Final Summary Report for Detailed Control Room Design Review of Nine Mile Point Unit One, July 1, 1985. 13. NRC report to NMPC, July 1, 1986, Safety Evaluation - Detailed Control Room Design Review (TAC 56141), NMP1. 14. NMPC letter to NRC, January 30, 1987, Supplemental DCRDR Summary Report. 15. Nine Mile Point Unit 1 Emergency Operating Procedure (Rev. 0) Flowcharts, Follow-on Task Analysis and Human Factors Review, December 1988, ARD Corporation. 16. Nine Mile Point Unit 1 EOP-4.1, SOP-1, and Operating Procedures Required by EOPs, Follow-on Task Analysis and Human Factors Review, January 1989, ARD Corporation. 17. NRC letter to NMPC, August 1990, Safety Evaluation for the Nine Mile Point Nuclear Station Unit 1 Detailed Control Room Design Review. 18. NMPC, MDC-4, Control Room Human Factors Design Criteria, Revision 4. B. SAFETY PARAMETER DISPLAY SYSTEM 1.0 Introduction to the Safety Parameter Display System As a result of the Three Mile Island incident in 1979, the NRC issued regulatory requirements regarding the improvement of plant monitoring systems at nuclear power facilities. One requirement dealt with the implementation of a safety parameter display system (SPDS) as described by NUREG-0660, May 1980, and NMP Unit 1 UFSAR Section XVIII XVIII-10 Rev. 25, October 2017 NUREG-0737, November 1980, as supplemented by Generic Letter 82-33, December 17, 1982. Additional information regarding the SPDS was given by NUREG-0800 Section 18.2, November 1984, and NUREG-1342, April 1989. The purpose of the SPDS is to continuously display information from which plant safety status can be readily and reliably assessed. The principal function of the SPDS is to aid control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by Operators to avoid a degraded core. In response to the SPDS guidance issued by the NRC, Unit 1 submitted its proposed SPDS design by letter dated January 3, 1984. The NRC conducted a review of the Unit 1 SPDS and a Safety Evaluation Report was issued on May 8, 1986. Unit 1 replied by letter on September 29, 1986, providing information requested by the NRC for them to complete their review. The NRC issued Generic Letter 89-06 on April 12, 1989, requesting all licensees to update the implementation status of the SPDS. A response to the NRC was provided on July 11, 1989, with certification that Unit 1 SPDS met the requirements of NUREG-0737, Supplement 1. A letter from the NRC on August 8, 1990, concluded that Unit 1 satisfactorily met all the requirements for a SPDS and considered implementation complete. 2.0 System Description The SPDS is a computer-based data acquisition and display subsystem of the Honeywell 4400 plant process computer. The computer and associated display system are non-Class 1E equipment. As such, they have not been seismically qualified nor have they been designed to meet single failure criterion. The system design includes isolation devices at interfaces between safety-related and nonsafety-related circuits to reduce the potential for the computer to adversely affect safety-related equipment. The SPDS display is available on two computer terminals in the control room, and a terminal each in the technical support center (TSC) and the emergency operations facility (EOF). System descriptions are included in the January 3, 1984, January 20, 1986, and September 29, 1986, Unit 1 submittals to the NRC. The SPDS consists of two major levels of display, the primary display, which includes a safety function status display, and NMP Unit 1 UFSAR Section XVIII XVIII-11 Rev. 25, October 2017 five secondary displays. There is a separate secondary display for each of the critical safety functions: reactivity control, core cooling, reactor vessel integrity, primary containment integrity, and radioactive release control. SPDS displays are accessed through a menu on the plant computer, and each of the SPDS displays has a unique display number that is entered to "call up" that display. The SPDS displays a concise, predetermined set of critical plant variables, defined as SPDS parameters, that are sufficient to provide the minimum information required to assess the critical safety functions. Generally, the analog SPDS parameters are shown as bar graphs on the associated safety function displays, with color coding used to indicate the status of the variable. The exact digital value and the direction and rate of change of each parameter are also displayed. At the bottom of each SPDS display, including the overview display, are five safety status blocks, one for each of the five critical safety functions. Each block is color coded to indicate the safety function status. Thus, the five critical safety functions can be evaluated from any SPDS display page. The status of a critical safety function block is derived from the status of the individual parameters associated with that block. 3.0 Role of the SPDS The SPDS continuously displays the safety function status of the plant during normal, abnormal, and emergency operations, plant status during stressful conditions, determine rapidly and reliably the safety status of the plant, and intervene in situations that require Operator action. 4.0 Human Factors Engineering Guidelines The SPDS is designed within the guidelines of NUREG-0700, particularly as they apply to cathode ray tube (CRT) displays. The human factors aspects of the SPDS design were evaluated during the DCRDR. The human factors manual for future design change provides additional basis for the analysis of human factors considerations of the information provided to the control room Operator. 5.0 Human Factors Engineering Principles Applied to the SPDS Design NMP Unit 1 UFSAR Section XVIII XVIII-12 Rev. 25, October 2017 NUREG-0737 Supplement 1 applies specific human factors engineering principles to the SPDS. The following subsections address the human factors engineering principles as applied to the Unit 1 SPDS. 5.1 NUREG-0737, Supplement 1, Section 4.1.a 5.1.1 Concise Display The critical set of plant variables is continuously presented on a CRT display in the control room. Secondary displays are available to provide more detailed information for each of the five safety functions. 5.1.2 Critical Plant Variables The SPDS displays a concise, predetermined set of critical plant variables that are sufficient to provide the minimum information required to assess the critical safety functions. Parameter selection evolved primarily through a plant-specific assessment of the applicability of the parameters presented on the generic BWROG SPDS displays, with additional consideration given to information contained in the following: 1. NSAC 21, "Fundamental Safety Parameter Set for Boiling Water Reactors." 2. A draft of the PSTGs that was developed from the generic BWROG EPGs. 3. Plant-specific Regulatory Guide (RG) 1.97 Type B and C variables. 4. Information obtained from discussion with Unit 1 plant operation and training personnel. A list of variables displayed on SPDS is given in Table XVIII-1. 5.1.3 Rapid and Reliable Determination of Safety Status The Unit 1 SPDS presents near real-time data to the control room Operator. The SPDS update rate is 5 sec. The sampling rate is variable based on the fastest scan rate of input points. Alarm checking is 60 sec for all SPDS output points. The SPDS responds to Operator commands in approximately 10 sec or less. Perceptual aids (color changes) inform the Operator of valid, NMP Unit 1 UFSAR Section XVIII XVIII-13 Rev. 25, October 2017 questionable and invalid status supported by the plant process computer and the SPDS subsystem software validity checking. 5.1.4 Aid to Control Room Personnel The Unit 1 SPDS is operated during normal, abnormal and emergency conditions. Critical plant variables are displayed on bar graphs that show the full range of the parameter values. The SPDS provides perceptual cues to alert the Operator to abnormal and emergency operating conditions. The bar graphs and safety status indicators change color to indicate the status of the parameter: green - normal condition yellow - warning condition red - alarm condition violet - not validated input signal 5.2 NUREG-0737, Supplement 1, Section 4.1.b 5.2.1 Convenient Location There are three CRT displays, centrally located in the control room adjacent to the E panel, which can display SPDS. This is convenient and readily accessible to the control room Operators. The SPDS displays do not interfere with the control room operating systems, or with displays important for safe operation. 5.2.2 Continuous Display One of the computer terminals in the control room is dedicated to the SPDS mode at all times. This is administratively controlled by the SPDS operating procedure. The safety function status blocks at the bottom of each SPDS display page permit all five critical safety functions to be evaluated on a continuous basis. The secondary display for a particular safety function can be selected when detailed information about that safety function is required. The SPDS is operable when the plant process computer is operable. The Technical Specifications specify compensatory measures to be taken when the process computer is not available. 5.3 NUREG-0737, Supplement 1, Section 4.1.c NMP Unit 1 UFSAR Section XVIII XVIII-14 Rev. 25, October 2017 5.3.1 Procedures and Training The Operators are trained on the SPDS and the EOPs together during normal requalification training. Simulator training includes training in the use of SPDS. 5.3.2 Isolation of SPDS from Safety-Related Systems The SPDS does not initiate any protective functions. The SPDS is electrically isolated from the safety-related sensing devices in the plant so that a failure of the SPDS will not affect the operation of the plant. Qualified isolation devices such as Foxboro Spec. 200 models N-2AO-VAI and N-2AI-T2V dual input and output modules isolation devices are used. Both modules are transformer isolated and qualified to IEEE-344-1975 and IEEE-323-1974. 5.4 NUREG-0737, Supplement 1, Section 4.1.e 5.4.1 Incorporation of Accepted Human Factors Engineering Principles The SPDS was reviewed by a human factors consultant as part of the DCRDR and several HEOs were identified regarding the human factors considerations incorporated into the design. Human factors related recommendations were incorporated into the SPDS design to correct the problems identified by the HEOs. Human factors reviews made by the human factors consultant and plant Operators demonstrated that the characteristics of the SPDS displays and other operational interfaces are sufficient to allow reasonable assurance that the information provided will be readily perceived and comprehended by its users. ARD Corporation performed the human factors review of SPDS, during the DCRDR, and has reviewed the modified SPDS. Specific human factors engineering principles added to the SPDS are listed below: 1. Display Color Coding The human factors review of the computer displays incorporated the NUREG-0700 CRT color recommendations into the SPDS displays. Specific recommendations for the application of these colors to each of the SPDS screens were made. The implementation of these recommendations on each of the individual display NMP Unit 1 UFSAR Section XVIII XVIII-15 Rev. 25, October 2017 screens was verified and found to be acceptable by ARD Corporation human factors specialists. 2. Display Format Recommendations were made to standardize the format of the SPDS displays. This was done to have information such as display titles in the same place on all displays and to have a consistent use of headings and demarcation lines. Specific recommendations were made for individual display screens. The implementation of each of these modifications was verified and found to be acceptable by the human factors reviewer. 5.4.2 Information Can Be Readily Perceived and Comprehended The SPDS has incorporated accepted human factors engineering principles using pattern and color coding to aide the Operator to detect and recognize unsafe operating conditions. Red, yellow, green, black, white, cyan, and violet are used in accordance with the guidance of NUREG-0700. All variables displayed on SPDS are identified by the variable description and the proper engineering units. 5.5 NUREG-0737, Supplement 1, Section 4.1.f, Sufficient Information The information provided to plant Operators by the SPDS is sufficient for the plant Operators to evaluate the following critical safety functions: reactivity control, reactor core cooling and heat removal from the primary system, reactor coolant system integrity, containment integrity, and radioactivity control. The SPDS parameter set is listed in Table XVIII-1. 6.0 Procedures 6.1 Operating Procedures Operating procedures ensure that at least one computer terminal in the control room is committed to the SPDS mode at all times. 6.2 Surveillance Procedures No specific tests or inspections are required for the SPDS. Inputs to the system are calibrated during normal plant tests of the sensing device. NMP Unit 1 UFSAR Section XVIII XVIII-16 Rev. 25, October 2017 The displays for SPDS are verified during normal instrumentation and control (I&C) surveillance procedures. The SPDS computer points are included in routine instrument loop surveillance procedures. Site procedures meet the requirement of NUREG-0737, Supplement 1, taking into account information provided in NUREG-1342. 7.0 References 1. NUREG-0696, Functional Criteria for Emergency Response Facilities, Section 5.0. 2. NUREG-0737, Supplement 1, Requirements for Emergency Response Capability, Section 4.0, Generic Letter 82-33. 3. NUREG-0800, Standard Review Plan, Section 18.2, Safety Parameter Display System, Rev. 0. 4. NUREG-1342, A Status Report Regarding Industry Implementation of Safety Parameter Display Systems. 5. NUREG-0700, Guidelines for Control Room Design Review. 6. Task Action Plan Item I.D.2-SPDS-10CFR50.54(f) - Generic Letter 89-06. 7. Nine Mile Point Nuclear Station Unit 1 SPDS Operating Performance Validation Summary Report (OEI 8309-16), Operations Engineering Inc. 8. NMPC letter to NRC, January 3, 1984, Submittal Regarding the SPDS. 9. Final Summary Report for Detailed Control Room Design Review of Nine Mile Point Unit One, July 1, 1985. 10. NMPC letter NMP1L 0019 to NRC, January 20, 1986, SPDS Final Design Configuration Nine Mile Point Unit 1. 11. NRC letter to NMPC, May 8, 1986, Safety Evaluation - Safety Parameter Display System for Nine Mile Point 1. 12. NMPC letter to NRC, NMP1L 0099, September 29, 1986, Response to Safety Evaluation of the Nine Mile Point Unit 1 SPDS. NMP Unit 1 UFSAR Section XVIII XVIII-17 Rev. 25, October 2017 13. NMPC letter NMP1L 0419 to NRC, July 11, 1989, Response to Generic Letter 89-06, Task Action Plan Item I.D.2-SPDS. 14. NRC letter to NMPC, August 8, 1990, Response to NRC Generic Letter 89-06 on the Safety Parameter Display System for Nine Mile Point Nuclear Station, Unit 1. NMP Unit 1 UFSAR Section XVIII XVIII-18 Rev. 25, October 2017 TABLE XVIII-1 SPDS PARAMETER SET KEY SPDS Display SFS - Safety Function Status Display SPDS Display PWR - Reactivity Control Display SPDS Display CLG - Core Cooling SPDS Display RCS - Reactor Vessel Integrity SPDS Display CNT - Primary Containment Integrity SPDS Display RRC - Radioactive Release Integrity Parameter SFS PWR CLG RCS CNT RRC APRM Reactor Power X X Containment Radiation Monitor X Control Rod Position X Drywell Floor Sump Rate X Containment Oxygen Concentration X Drywell Pressure X X X Drywell Temperature X IRM Position X Main Stack Activity X X MSIV Position X Offgas Dose Rate X Reactor Recirc Flow X RPV Pressure X X RPV Water Level X X Safety/ERV Position X SRM Count Rate X SRM Position X Torus Level X Torus Temperature X X Containment Continuous Air Monitor X IRM Reactor Power X

NMP Unit 1 UFSAR APPENDIX A Appendix A A-i Rev. 25, October 2017 This Appendix was not used.

NMP Unit 1 UFSAR Appendix B B-i Rev. 25, October 2017 TABLE OF CONTENTS Section Title APPENDIX B NINE MILE POINT NUCLEAR STATION, LLC, QUALITY ASSURANCE PROGRAM TOPICAL REPORT, NINE MILE POINT NUCLEAR STATION UNITS 1 AND 2 OPERATIONS PHASE NMP Unit 1 UFSAR Appendix B B-1 Rev. 25, October 2017 APPENDIX B NINE MILE POINT NUCLEAR STATION, LLC QUALITY ASSURANCE PROGRAM TOPICAL REPORT NINE MILE POINT NUCLEAR STATION UNITS 1 AND 2 OPERATIONS PHASE The previous Nine Mile Point Nuclear Station, LLC, Quality Assurance Program Topical Report, Nine Mile Point Nuclear Station Units 1 and 2, Operations Phase, has been superseded by Constellation Generation Group, LLC, Quality Assurance Topical Report (QATR), approved by the Nuclear Regulatory Commission (NRC) on December 21, 2006. The previous Appendix B has been deleted. The effective QATR is maintained as a separate document.

NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Appendix C C-i Rev. 25, October 2017 APPENDIX C LICENSE RENEWAL SUPPLEMENT - AGING MANAGEMENT PROGRAMS AND TIME-LIMITED AGING ANALYSES C.0 INTRODUCTION C.1 AGING MANAGEMENT PROGRAMS C.1.1 10CFR50 Appendix J Program C.1.2 ASME Section XI Inservice Inspection (Subsection IWE) Program C.1.3 ASME Section XI Inservice Inspection (Subsection IWF) Program C.1.4 ASME Section XI Inservice Inspection (Subsections IWB, IWC, IWD) Program C.1.5 Boraflex Monitoring Program C.1.6 Buried Piping and Tanks Inspection Program C.1.7 BWR Feedwater Nozzle Program C.1.8 BWR Penetrations Program C.1.9 BWR Reactor Water Cleanup System Program C.1.10 BWR Stress Corrosion Cracking Program C.1.11 BWR Vessel ID Attachment Welds Program C.1.12 BWR Vessel Internals Program C.1.13 Closed-Cycle Cooling Water System Program C.1.14 Compressed Air Monitoring Program C.1.15 Environmental Qualification Program C.1.16 Fatigue Monitoring Program C.1.17 Fire Protection Program C.1.18 Fire Water System Program C.1.19 Flow-Accelerated Corrosion Program C.1.20 Fuel Oil Chemistry Program C.1.21 Fuse Holder Inspection Program C.1.22 Inspection of Overhead Heavy Load and Light Load Handling Systems Program C.1.23 Masonry Wall Program C.1.24 Non-EQ Electrical Cables and Connections Program C.1.25 Non-EQ Electrical Cables and NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Appendix C C-ii Rev. 25, October 2017 Connections Used in Instrumentation Circuits Program C.1.26 Non-Segregated Bus Inspection Program C.1.27 One-Time Inspection Program C.1.28 Open-Cycle Cooling Water System Program C.1.29 Preventive Maintenance Program C.1.30 Reactor Head Closure Studs Program C.1.31 Reactor Vessel Surveillance Program C.1.32 Selective Leaching of Materials Program C.1.33 Structures Monitoring Program C.1.34 Systems Walkdown Program C.1.35 Torus Corrosion Monitoring Program C.1.36 Water Chemistry Control Program C.1.37 Bolting Integrity Program C.1.38 BWR Control Rod Drive Return Line Nozzle Program C.1.39 Protective Coating Monitoring and Maintenance Program C.1.40 Non-EQ Electrical Cable Metallic Connections Inspection Program C.1.41 Drywell Supplemental Inspection Program C.2 TIME-LIMITED AGING ANALYSES SUMMARIES C.2.1 Reactor Vessel Neutron Embrittlement Analysis C.2.1.1 Upper-Shelf Energy C.2.1.2 Pressure-Temperature Limits C.2.1.3 Elimination of Circumferential Weld Inspection C.2.1.4 Axial Weld Failure Probability C.2.2 Metal Fatigue Analysis C.2.2.1 Reactor Vessel Fatigue Analysis C.2.2.2 Feedwater Nozzle and Control Rod Drive Return Line Nozzle Fatigue and Cracking Analyses C.2.2.3 Non-ASME Section III Class 1 Piping and Components Fatigue Analysis C.2.2.4 Reactor Vessel Internals Fatigue Analysis NMP Unit 1 UFSAR TABLE OF CONTENTS Section Title Appendix C C-iii Rev. 25, October 2017 C.2.2.5 Environmentally-Assisted Fatigue C.2.2.6 Fatigue of the Emergency Condenser C.2.3 Environmental Qualification C.2.3.1 Electrical Equipment EQ C.2.4 Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis C.2.4.1 Torus Shell and Vent System Fatigue Analysis C.2.4.2 Torus-Attached Piping Analysis C.2.4.3 Torus Wall Thickness C.2.4.4 Fatigue of Primary Containment Penetrations C.2.5 Other Plant-Specific TLAAs C.2.5.1 Reactor Vessel and Reactor Vessel Closure Head Weld Flaw Evaluations C.2.5.2 Reactor Water Cleanup System Weld Overlay Fatigue Flaw Growth Evaluations C.3 GENERIC QUALITY ASSURANCE PROGRAM REQUIREMENTS FOR LICENSE RENEWAL C.4 REFERENCES NMP Unit 1 UFSAR Appendix C C-iv Rev. 25, October 2017 LIST OF TABLES Table Number Title C-1 COMMITMENTS NMP Unit 1 UFSAR Appendix C C-1 Rev. 25, October 2017 APPENDIX C LICENSE RENEWAL SUPPLEMENT - AGING MANAGEMENT PROGRAMS AND TIME-LIMITED AGING ANALYSES C.0 INTRODUCTION The original operating license for Nine Mile Point Nuclear Station - Unit 1 (Unit 1) was issued by the Nuclear Regulatory Commission (NRC) on August 22, 1969, and authorized operation for 40 yr. Per 10CFR54, licensees could apply for a renewed operating license that would authorize up to an additional 20 yr of operation. Unit 1 applied for a renewed license on May 26, 2004, and amended the application on July 14, 2005. The NRC granted approval of a renewed license on October 31, 2006. The environmental and safety reviews conducted by the NRC are documented in NUREG-1437 Supplement 24, and NUREG-1900, respectively. This appendix to the Unit 1 Updated Final Safety Analysis Report (UFSAR) meets the requirements of 10CFR54.21(d) and describes the programs credited for managing the aging of applicable structures, systems and components (SSC), describes the time-limited aging analyses (TLAA) performed for license renewal, lists the commitments made to meet the regulations, and describes the generic quality assurance program requirements for license renewal. The Aging Management Programs (AMP) described in this appendix have been credited with managing the aging of SSCs that have been determined to be within the scope of license renewal [per 10CFR54.4(a)] and subject to aging management [per 10CFR54.21(a)(1)]. There are 41 AMPs described in this appendix. Six are new programs while the remaining 35 are existing (some have license renewal names but they include activities already performed at the station). For each AMP, a description of the program is provided along with the identification of any enhancements (i.e., commitments) and/or exceptions taken to the NRC guidance documents (NUREG-1800 Rev. 0 and NUREG-1801 Rev. 0). This appendix also describes the TLAA dispositions performed for license renewal per 10CFR54.21(c). The dispositions address existing Unit 1 calculations and analyses that include, as one of the criteria, a time limit. The time limit is normally the duration of the original license, i.e., 40 yr, but can be any NMP Unit 1 UFSAR Appendix C C-2 Rev. 25, October 2017 length of time. Each TLAA description includes the scope of the evaluation and a conclusion of how the evaluation is dispositioned for the period of extended operation (i.e., 41 to 60 yr). A table documenting each of the commitments made as part of the license renewal application (LRA) is also provided. Each commitment is associated with a TLAA or AMP and is described in those sections. The table provides a single location for all the commitments. The source document for the commitment is listed along with the due date. Where the source document is listed as "LRA Section...," it refers to the Amended License Renewal Application (ALRA) submitted on July 14, 2005, under letter number NMP1L 1962. A due date of "Prior to period of extended operation" means prior to August 22, 2009. The final section of this appendix addresses the generic quality assurance program requirements for license renewal. Each AMP must meet three attributes that are the same for all AMPs. The attributes are corrective actions, confirmation process, and administrative controls. The final section describes how each attribute is addressed for the Unit 1 AMPs. C.1 AGING MANAGEMENT PROGRAMS C.1.1 10CFR50 Appendix J Program The 10CFR50 Appendix J Program detects degradation of the containment structure and components that comprise the containment pressure boundary, including seals and gaskets. Containment leak rate tests are performed to assure that leakage through the primary containment and systems and components penetrating primary containment does not exceed allowable leakage limits specified in the Technical Specifications. This program complies with Option B requirements of 10CFR50 Appendix J, with plant-specific exceptions approved by the NRC as part of license amendments, and implements the guidelines provided in NRC Regulatory Guide (RG) 1.163 and NEI 94-01. C.1.2 ASME Section XI Inservice Inspection (Subsection IWE) Program The American Society of Mechanical Engineers (ASME) Section XI Inservice Inspection (Subsection IWE) Program (referred to herein as the IWE ISI Program) manages aging effects due to 1) corrosion of carbon steel components comprising the containment pressure boundary; and 2) degradation of containment NMP Unit 1 UFSAR Appendix C C-3 Rev. 25, October 2017 pressure-retaining polymers. Program activities include visual examination, with limited surface or volumetric examinations when augmented examination is required. The IWE ISI Program is based on the 1998 edition of the ASME Boiler and Pressure Vessel Code, Section XI (Subsection IWE), for containment inservice inspection with plant-specific exceptions approved by the NRC. This is an exception to the evaluation in NUREG-1801 (which covers ASME Section XI requirements from both the 1992 edition with the 1992 addenda, and the 1995 edition with the 1996 addenda). The Unit 1 ASME Section XI Inservice Inspection (Subsection IWE) Program is being improved to add an augmented VT-1 visual examination of the Unit 1 containment penetration bellows. This inspection will be performed using enhanced techniques qualified for detecting stress corrosion cracking (SCC) per NUREG-1611, Table 2, Item 12. This improvement is not required for consistency with NUREG-1801 but is an activity being adopted to ensure consistency with industry practice. C.1.3 ASME Section XI Inservice Inspection (Subsection IWF) Program The ASME Section XI Inservice Inspection (Subsection IWF) Program (referred to herein as the IWF ISI Program) manages aging of carbon steel component and piping supports, including ASME Class MC supports, due to general corrosion and wear. Program activities include visual examination to determine the general mechanical and structural condition of components and their supports. The IWF ISI Program is based on the 1989 edition of the ASME Boiler and Pressure Vessel Code, Section XI (Subsection IWF), for inservice inspection of supports, and implements the alternate examination requirements of ASME Code Case N-491-1. There are exceptions to the evaluation in NUREG-1801 (which covers ASME Section XI requirements from the 1989 edition through the 1995 edition and addenda through the 1996 addenda). C.1.4 ASME Section XI Inservice Inspection (Subsections IWB, IWC, IWD) Program The ASME Section XI Inservice Inspection (Subsections IWB, IWC, IWD) Program manages aging of Class 1, 2, or 3 pressure-retaining components and their integral attachments. Program activities include periodic visual, surface, and/or volumetric examination and pressure tests of Class 1, 2, and 3 pressure-retaining components. The ASME Section XI Inservice NMP Unit 1 UFSAR Appendix C C-4 Rev. 25, October 2017 Inspection (Subsections IWB, IWC, IWD) Program is based on ASME Section XI, 1989 edition with no addenda, and ASME Section XI, Appendix VIII, 1995 edition through 1996 addenda. Examination categories B-F, B-J, C-F-1, C-F-2, and intergranular stress corrosion cracking (IGSCC) Category A are inspected using NRC-approved risk-informed methodology. Prior to the period of extended operation, the ISI Program will be updated to the latest edition and addenda of ASME Section XI, as mandated by 10CFR50.55a and 10CFR54 requirements. There are exceptions to the program described in NUREG-1801 (which cites ASME Section XI requirements covered in the 1995 edition through 1996 addenda). C.1.5 Boraflex Monitoring Program The Boraflex Monitoring Program manages degradation of neutron absorbing material in spent fuel pool storage racks resulting from radiation exposure and possible water ingress. Program activities include 1) inspection of the test coupons to detect dimensional changes; 2) correlation of measured levels of silica in the spent fuel pool with analysis using a predictive code (e.g., RACKLIFE) to estimate boron loss from Boraflex panels; and 3) neutron attenuation testing to measure the boron areal density of the short-length test coupons. The Boraflex Monitoring Program is based on existing technology and methods for testing and evaluating material properties necessary to ensure the required 5-percent margin to criticality in the spent fuel pool is maintained. The Boraflex Monitoring Program for Unit 1 will be enhanced to perform periodic in-situ neutron attenuation testing and measurement of boron areal density for those Boraflex racks that remain in use during the period of extended operation. It will also be enhanced to create a new activity which provides instruction for the trending of silica levels, coupon results, and the results of in-situ testing. Enhancements will be completed prior to the period of extended operation. C.1.6 Buried Piping and Tanks Inspection Program The Buried Piping and Tanks Inspection Program is a new program that will manage the aging effects on the external surfaces of carbon steel, low-alloy steel, and cast iron components (e.g., tanks, piping) that are buried in soil. Program activities will include visual inspections of external coatings and wrappings to detect damage and degradation. Prior to entering the period of extended operation, Nine Mile Point will verify that there has NMP Unit 1 UFSAR Appendix C C-5 Rev. 25, October 2017 been at least one opportunistic or focused inspection within the past 10 yr. Upon entering the period of extended operation, Nine Mile Point will perform a focused inspection within 10 yr, unless an opportunistic inspection occurred within this 10-yr period. All credited inspections will be performed in areas with the highest likelihood of corrosion problems, and in areas with a history of corrosion problems. This program will be implemented prior to the period of extended operation. C.1.7 BWR Feedwater Nozzle Program The Unit 1 Feedwater Nozzle Program requires ultrasonic test (UT) inspections of the feedwater nozzles every 10 yr to verify the nozzles are acceptable for continued service. The Feedwater Nozzle Program is implemented through the ISI Program which, at the time the LRA was submitted, conformed to the requirements in ASME Section XI, Subsection IWB, Table IWB 2500-1 (1989 edition, no addenda), and ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," 1995 edition through 1996 addenda. NUREG-1801, Section XI.M5, identifies the 1995 edition (including the 1996 addenda) of ASME Section XI as the basis for the Generic Aging Lessons Learned (GALL) Feedwater Nozzle Program. The ISI Programs will not comply with the edition and addenda of ASME Section XI cited in the GALL report because the programs are updated to the latest edition and addenda of ASME Section XI, as mandated by 10CFR50.55a, prior to the start of each inspection interval. UT and particle test inspections required by NUREG-0619 have been superseded because the inspections are now performed in accordance with ASME Section XI, Appendix VIII. C.1.8 BWR Penetrations Program The Boiling Water Reactor (BWR) Penetrations Program manages the effects of cracking in the various penetrations of the reactor pressure vessels (RPV) at Nine Mile Point. The BWR Penetrations Program is based on guidelines issued by the BWR Vessel and Internals Project and approved by the NRC. This program is implemented by the BWR Vessel and Internals Program (BWRVIP) for managing specific aging effects. The attributes of the BWR Penetrations Program related to maintaining reactor coolant water chemistry are included in the Water Chemistry Control Program. NMP Unit 1 UFSAR Appendix C C-6 Rev. 25, October 2017 C.1.9 BWR Reactor Water Cleanup System Program The BWR Reactor Water Cleanup (RWCU) System Program manages the effects of stress corrosion cracking or IGSCC on the intended function of austenitic stainless steel piping in the RWCU system. This program is based on the NRC criteria related to inspection guidelines for RWCU piping welds outboard of the second isolation valve, as delineated in NUREG-0313, Revision 2, and Generic Letter (GL) 88-01. An exception is taken to the acceptance criteria program element in that Unit 1 utilizes the 1989 edition with no addenda of the ASME Section XI Code versus the 1995 edition through the 1996 addenda as defined in the GALL report. The attributes of the BWR RWCU System Program related to maintaining reactor coolant water chemistry are included in the Water Chemistry Control Program. C.1.10 BWR Stress Corrosion Cracking Program The BWR Stress Corrosion Cracking (SCC) Program manages IGSCC in reactor coolant pressure boundary (RCPB) piping made of stainless steel, as delineated in NUREG-0313, Revision 2, and GL 88-01 and its Supplement 1, as modified by BWRVIP-75. Augmented inspections are performed in accordance with these documents. An exception to the program described in NUREG-1801 is that the acceptance criteria for the BWR SCC Program are based upon the 1989 edition of the ASME Section XI Code versus the 1995 edition through the 1996 addenda, as described in NUREG-1801. The attributes of the BWR SCC Program related to maintaining reactor coolant water chemistry are included in the Water Chemistry Control Program. C.1.11 BWR Vessel ID Attachment Welds Program The BWR Vessel ID Attachment Welds Program manages the effects of cracking in RPV inside diameter (ID) attachment welds. This program is based on industry guidelines issued by the BWRVIP and approved by the NRC. The BWR Vessel ID Attachment Welds Program is implemented by the BWRVIP for managing specific aging effects. The attributes of the BWR Vessel ID Attachment Welds Program related to maintaining reactor coolant water chemistry are included in the Water Chemistry Control Program. C.1.12 BWR Vessel Internals Program The BWRVIP manages aging of materials inside the reactor vessel. Program activities include 1) inspections for the presence and effects of cracking; and 2) monitoring and control of water NMP Unit 1 UFSAR Appendix C C-7 Rev. 25, October 2017 chemistry. This program is based on guidelines issued by the BWRVIP and approved (or pending approval) by the NRC. Inspections and evaluations of reactor vessel components are consistent with the guidelines provided in the following BWRVIP reports: BWRVIP-18, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines BWRVIP-25, BWR Core Plate Inspection and Flaw Evaluation Guidelines BWRVIP-26, BWR Top Guide Inspection and Flaw Evaluation Guidelines BWRVIP-27, BWR Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines BWRVIP-38, BWR Shroud Support Inspection and Flaw Evaluation Guidelines BWRVIP-47, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines BWRVIP-48, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines BWRVIP-49, Instrument Penetration Inspection and Flaw Evaluation Guidelines BWRVIP-74, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines BWRVIP-76, BWR Core Shroud Inspection and Flaw Evaluation Guidelines Unit 1 has completed, or will complete, each of the license renewal applicant action items described in the NRC safety evaluations for these BWRVIP reports. In addition, Unit 1 will implement the NRC-approved inspection and flaw evaluation guidelines for the steam dryer and inaccessible core spray component welds when issued. The attributes of the BWRVIP related to maintaining reactor coolant water chemistry are included in the Water Chemistry Control Program. Enhancements to the BWRVIP include the following revisions to existing activities that are credited for license renewal: 1. The reinspection scope and frequency for the grid beam going forward will be based on BWRVIP-26A guidance for plant-specific flaw analysis and crack growth NMP Unit 1 UFSAR Appendix C C-8 Rev. 25, October 2017 assessment. The maximum reinspection interval for the grid beam will not exceed 10 yr, consistent with standard BWRVIP guidance for the core shroud. The reinspection scope will be equivalent to the UT baseline 2005 inspection scope. In addition, the reinspection scope will include an EVT-1 sample inspection of at least two locations with accessible indications within the initial 6 yr of the 10-yr interval. The intent of the EVT-1 is to monitor the known cracking to confirm flaw analysis crack growth assumptions. 2. Nine Mile Point will implement the resolution of the open item documented in BWRVIP-18 regarding the inspection of inaccessible welds for core spray. It will be included in the BWRVIP response to be reviewed and accepted by the NRC. 3. Once the guidelines for inspection and evaluation for steam dryers currently under development by the BWRVIP committee are documented, reviewed and accepted by the NRC, the actions will be implemented in accordance with the BWRVIP. 4. The baseline inspections recommended in BWRVIP-47 for the BWR lower plenum components will be incorporated into the program. 5. If the October 19, 2005, draft of Code Case N-730 is approved by ASME, Unit 1 will implement the final Code case as conditioned by the NRC. If the Code case is approved by ASME but not yet listed in RG 1.147, Unit 1 will seek NRC approval of the Code case on a plant-specific basis as conditioned by the NRC. It will be programmatically required that during the period of extended operation, should a control rod drive (CRD) stub tube rolled in accordance with the provisions of the Code case resume leaking, Unit 1 will implement one of the following zero leakage permanent repair strategies prior to startup from the outage in which the leakage was detected: a. A welded repair consistent with BWRVIP-58-A, "BWRVIP Internal Access Weld Repair" and Code Case N-606-1, as endorsed by the NRC in RG 1.147. NMP Unit 1 UFSAR Appendix C C-9 Rev. 25, October 2017 b. A variation of the welded repair geometry specified in BWRVIP-58-A subject to the approval of the NRC using Code Case N-606-1. c. A future developed mechanical/welded repair method subject to the approval of the NRC. 6. Unit 1 will evaluate component susceptibility to loss of fracture toughness due to neutron fluence and thermal embrittlement. Assessments and inspections will be performed, as necessary, to ensure that intended functions are not impacted by the aging effect. 7. An EVT-1 examination of the Unit 1 feedwater sparger end bracket welds will be added to the BWRVIP. The inspection extent and frequency of the end bracket weld inspection will be the same as the ASME Section XI inspection of the feedwater sparger bracket vessel attachment welds. 8. Unit 1 will perform an EVT-1 inspection of the thermal shield to flow shield weld starting in 2007, and proceeding at a 10-yr frequency thereafter consistent with the ISI inspection interval. Enhancements will be completed prior to the period of extended operation. C.1.13 Closed-Cycle Cooling Water System Program The Closed-Cycle Cooling Water System (CCCWS) Program manages loss of material and fouling of components exposed to CCCW environments. The applicable piping systems include the reactor building closed loop cooling (RBCLC) system, control room heating, ventilation and air conditioning (HVAC) system, and the heat exchanger jacket water cooling portions of the emergency diesel generator (EDG) system. Also included are portions of non-safety related systems credited in the aging management review. Program activities include chemistry monitoring, surveillance testing, data trending, and component inspections. The CCCWS Program implements the guidelines for controlling system performance and aging effects described in Electric Power Research Institute (EPRI) Report TR-107396. However, specific exception is taken to maintaining chemical corrosion inhibitor concentrations. NMPNS utilizes corrosion control without chemicals (i.e., pure water) for the RBCLC and control room HVAC NMP Unit 1 UFSAR Appendix C C-10 Rev. 25, October 2017 systems, and the EDG system chromate concentrations are maintained higher per EDG vendor recommendation. This is an exception to the program described in NUREG-1801. Enhancements to the CCCWS Program include the following revisions to existing activities that are credited for license renewal: 1. Direct periodic inspections to monitor for loss of material in the piping of the CCCW systems. 2. Implement a Corrosion Monitoring Program for larger bore CCCW piping not subject to inspection under another program. 3. Establish periodic monitoring, trending, and evaluation of performance parameters for the RBCLC and control room HVAC systems. 4. Establish the frequencies to inspect for degradation of components in CCCW systems, including heat exchanger tube wall thinning. 5. Perform a heat removal capability test for the control room HVAC system at least every 5 yr. 6. Expand periodic chemistry checks of CCCW systems consistent with the guidelines of EPRI TR-107396. 7. Provide the controls and sampling necessary to maintain water chemistry parameters in CCCW systems within the guidelines of EPRI Report TR-107396. 8. Ensure acceptance criteria are specified in the implementing procedures for the applicable indications of degradation. The enhancements will be completed prior to the period of extended operation. C.1.14 Compressed Air Monitoring Program The Compressed Air Monitoring Program manages aging effects for portions of the compressed air systems within the scope of license renewal, including cracking and loss of material due to general corrosion, by controlling the internal environment of systems and components. Program activities include air quality NMP Unit 1 UFSAR Appendix C C-11 Rev. 25, October 2017 checks at various locations to detect contaminants that would affect the system's intended function. Additional visual inspections are credited for identification and monitoring of degradation for air compressors, receivers, and air dryers. The Compressed Air Monitoring Program is based on GL 88-14 and recommendations presented in Institute of Nuclear Power Operations (INPO) Significant Operating Event Report (SOER) 88-01. The program also includes good practice elements of the general maintenance and inspection activities for the compressor, receiver, and air drier discussed in EPRI TR-108147 (revision to EPRI NP-7079) and ASME OM-S/G-1998, Part 17. However, specific exception is taken to any maintenance recommended in EPRI TR-108147 that is not also endorsed by the equipment manufacturers, and to the preservice and inservice testing guidelines of ASME OM-S/G-1998, Part 17. This is an exception to the program described in NUREG-1801. Unit 1 also takes exception to the use of ISA-S7.0.01-1996 for air quality standards. The system air quality is monitored and maintained in compliance with the requirements of ANSI/ISA-S7.3-1975, which meets or exceeds the quality requirements for dew point, hydrocarbons, and particulate of Section 4.4 of EPRI TR-108147 and ISA-S7.0.01-1996. Enhancements to the Compressed Air Monitoring Program include the following revisions to existing activities that are credited for license renewal: 1. Develop new activities to manage the loss of material, stress corrosion cracking, and perform periodic system leak checks. 2. Expand the scope, periodicity, and inspection techniques to ensure that the aging of certain subcomponents of the dryers and compressors (e.g., valves, heat exchangers) is managed. 3. Establish activities that manage the aging of the internal surfaces of carbon steel piping and that require system leak checks to detect deterioration of the pressure boundaries. 4. Expand the acceptance criteria to ensure that the aging of certain subcomponents of the dryers and compressors (e.g., valves, heat exchangers) is managed. NMP Unit 1 UFSAR Appendix C C-12 Rev. 25, October 2017 5. Develop and implement the activities to address the failure mechanism of stress corrosion cracking in unannealed red brass piping in Unit 1. Enhancements will be completed prior to the period of extended operation. C.1.15 Environmental Qualification Program The Environmental Qualification (EQ) Program manages thermal, radiation, and cyclical aging for electrical equipment important to safety and located in harsh plant environments at Unit 1. Program activities 1) identify applicable equipment and environmental requirements; 2) establish, demonstrate, and document the level of qualification (including configuration, maintenance, surveillance, and replacement requirements); and 3) maintain (or preserve) qualification. The EQ Program employs aging evaluations based on 10CFR50.49(f) qualification methods. Components in the EQ Program must be refurbished, replaced, or have their qualification extended prior to reaching the aging limits established in the evaluation. Important attributes for the reanalysis of an aging evaluation include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria, and corrective actions (if acceptance criteria are not met). C.1.16 Fatigue Monitoring Program The Fatigue Monitoring Program (FMP) manages the fatigue life of RCPB components by tracking and evaluating key plant events. The FMP monitors operating transients to date, calculates cumulative usage factors (CUF) to date, and directs performance of engineering evaluations to develop preventive and mitigative measures in order not to exceed the design limit on fatigue usage. The FMP will be enhanced with guidance for the use of the FatiguePro software package and updated methodology for environmental fatigue factors in establishing updated fatigue life calculations for components, and to add safety relief valve (SRV) actuations for Unit 1 as a monitored transient. These enhancements will be completed prior to the period of extended operation. C.1.17 Fire Protection Program NMP Unit 1 UFSAR Appendix C C-13 Rev. 25, October 2017 The Fire Protection Program provides guidance for performance of periodic visual inspections to manage aging of the various materials comprising rated fire barriers. These include 1) sealants in rated penetration seals (subject to shrinkage due to weathering); 2) concrete and steel in fire-rated walls, ceilings, and floors (subject to loss of material due to flaking and abrasion; separation and concrete damage due to relative motion, vibration, and shrinkage); and 3) steel in rated fire doors (subject to loss of material due to corrosion and wear or mechanical damage). In addition, this program requires testing of the diesel-driven fire pump to verify that it is performing its intended function. This activity manages aging of the fuel oil supply line to, and the exhaust system from, the diesel engine, both of which may experience loss of material due to corrosion. Inspection and testing is performed in accordance with the guidance of applicable standards. There are two exceptions to the Fire Protection Program as described in NUREG-1801. Inspections on hollow metal fire doors will be performed on a plant-specific schedule, and valve lineups will not be used for aging management of fire suppression systems. These exceptions are consistent with NRC Interim Staff Guidance (ISG) 04. The Fire Protection Program will be enhanced to include the following: 1) periodic visual inspections of piping and fittings in a non-water environment in the Halon and carbon dioxide (CO2) fire suppression systems components to detect signs of degradation; 2) periodic functional tests of the diesel-driven fire pump will be enhanced to include inspection of engine exhaust system components to verify that loss of material is managed; and 3) the fire door inspection frequency will be determined by a plant-specific analysis. These enhancements will be completed prior to the period of extended operation. C.1.18 Fire Water System Program The Fire Water System Program manages aging of water-based fire protection systems due to loss of material and biofouling. Program activities include periodic maintenance, testing, and inspection of system piping and components containing water (e.g., sprinklers, nozzles, fittings, valves, hydrants, hose stations, standpipes). Inspection and testing is performed in accordance with the guidance of applicable National Fire Protection Association (NFPA) Codes and Standards and the Nuclear Electric Insurance Limited (NEIL) Members' Manual. NMP Unit 1 UFSAR Appendix C C-14 Rev. 25, October 2017 Enhancements to the Fire Water System Program include the following revisions to existing activities that are credited for license renewal: 1. Incorporate inspections to detect and manage loss of material due to corrosion into existing periodic test procedures. 2. Specify periodic component inspections to verify that loss of material is being managed. 3. Add procedural guidance for performing visual inspections to monitor internal corrosion and detect biofouling. 4. Develop new procedures and preventive maintenance tasks to implement sprinkler head replacement and/or inspections to meet NFPA 25, Section 5.3.1 (2003 edition) requirements. 5. Add requirements to periodically check the water-based fire protection systems for microbiological contamination. 6. Measure fire protection system piping wall thickness using non-intrusive techniques (e.g., volumetric testing) to detect loss of material due to corrosion. 7. Establish an appropriate means of recording, evaluating, reviewing, and trending the results of visual inspections and volumetric testing. 8. Define acceptance criteria for visual inspections and volumetric testing. Enhancements will be completed prior to the period of extended operation. C.1.19 Flow-Accelerated Corrosion Program The Flow-Accelerated Corrosion (FAC) Program (also referred to as the Erosion/Corrosion Program) manages aging effects due to FAC in carbon steel and low-alloy steel piping containing single-phase and two-phase high-energy fluids. Program activities include 1) analysis using a predictive code (CHECWORKS) to determine critical locations; 2) baseline NMP Unit 1 UFSAR Appendix C C-15 Rev. 25, October 2017 inspections to determine the extent of thinning at the selected locations; 3) follow-up inspections to confirm the predictions; and 4) repair or replacement of components, as necessary. The program considers the recommended actions in NRC Bulletin 87-01 and Information Notice 91-18, and implements the guidelines for an effective FAC Program presented in EPRI Report NSAC-202L-R2. The program also implements the recommendations provided in NRC Generic Letter (GL) 89-08, "Erosion/Corrosion Induced Pipe Wall Thinning." C.1.20 Fuel Oil Chemistry Program The Fuel Oil Chemistry Program manages loss of material due to corrosion that may result from introduction of contaminants into the plant's fuel oil tanks. Program activities include 1) sampling and chemical analysis of the fuel oil inventory at the plant; 2) sampling, testing, and analysis of new fuel oil as it is unloaded at the plant; and 3) cleaning and inspection of fuel oil tanks. The Fuel Oil Chemistry Program is based on maintaining fuel oil quality in accordance with the guidelines of American Society for Testing Materials (ASTM) Standards D975, D1796, D2276, and D4057. The Fuel Oil Chemistry Program takes exceptions to the following NUREG-1801, Section XI.M30 (Fuel Oil Chemistry Program) evaluation elements: 1. Unit 1 takes exception to using both ASTM D1796 and ASTM D2709 to determine the concentration of water and sediment in the diesel fuel oil tanks. Unit 1 uses only the guidance given in ASTM D1796. 2. Unit 1 takes exception to using the modified ASTM D2276, Method A, which specifies a pore size of 3.0 µm. Unit 1 uses a filter with a pore size of 0.8 µm as specified in ASTM D2276. 3. Unit 1 takes exception to multilevel sampling in the diesel fuel oil tanks. The physical configuration of the fuel oil tanks does not allow a representative fuel oil sample to be taken at multiple levels. 4. Unit 1 takes exception to periodically sampling the fuel oil day tanks. These small tanks do not have a provision for sampling. NMP Unit 1 UFSAR Appendix C C-16 Rev. 25, October 2017 5. Unit 1 takes exception to periodic internal inspection of any fuel oil day tank. The physical size and configuration are not suitable for such inspections and, after enhancement, all such tanks will be routinely drained, thereby removing any contaminants from the tank that would provide an aging mechanism. 6. Unit 1 takes exception to the addition of biocides, stabilizers, and corrosion inhibitors to the diesel fuel oil storage tanks. Enhancements to the Fuel Oil Chemistry Program include the following revisions to existing activities that are credited for license renewal: 1. Add a requirement for quarterly trending of water, sediment, and particulate contamination analysis results. 2. Add requirements to periodically inspect the interior surfaces of the emergency diesel generator fuel oil tanks for evidence of significant degradation, including a requirement that the tank bottom thickness be determined. Bottom thickness measurements will be performed using UT or other industry-recognized methods. 3. Provide guidelines for the appropriate use of biocides, corrosion inhibitors, and fuel stabilizers to maintain fuel oil quality. 4. Ensure acceptance criteria are specified in the implementing procedures for the applicable indications of potential degradation. 5. Add periodic opening of the diesel fire pump fuel oil day tank drain. 6. Add steps for removal of water, if found. Enhancements will be completed prior to the period of extended operation. C.1.21 Fuse Holder Inspection Program The Fuse Holder Inspection Program is a new plant-specific program that applies to fuse holders located outside of active NMP Unit 1 UFSAR Appendix C C-17 Rev. 25, October 2017 devices that have aging effects requiring management. This program requires testing to detect deterioration of metallic clamps that would affect the ability of in-scope fuse holders to perform their intended function. The Fuse Holder Inspection Program includes the following aging stressors: moisture, fatigue, ohmic heating, mechanical stress, vibration, thermal cycling, electrical transients, chemical contamination, oxidation, and corrosion. Analytical trending will not be included in this activity because the parameters monitored may vary depending upon the test method selected. This is an exception to the "Monitoring and Trending" element in Appendix A.1.2.3.5 to NUREG-1800, but is consistent with the latest regulatory and industry license renewal precedence. This program will be implemented prior to the period of extended operation. C.1.22 Inspection of Overhead Heavy Load and Light LoadHandling Systems Program The Inspection of Overhead Heavy Load and Light Load Handling Systems Program (referred to herein as the Crane Inspection Program) manages loss of material due to corrosion of cranes within scope of license renewal (WSLR). Program activities include 1) performance of various maintenance activities on a specified frequency; and 2) preoperational inspections of equipment prior to lifting activities. Crane inspection activities are based on the mandatory requirements of applicable industry standards and implement the guidance of NUREG-0612. The Crane Inspection Program will be enhanced to add specific direction for performance of corrosion inspections of certain hoist-lifting assembly components. The enhancement will be completed prior to the period of extended operation. C.1.23 Masonry Wall Program The Masonry Wall Program manages aging effects so that the evaluation basis established for each masonry wall WSLR remains valid through the period of extended operation. The Masonry Wall Program is based on the structures monitoring requirements of 10CFR50.65. The Masonry Wall Program is implemented by the Structures Monitoring Program for managing specific aging effects. C.1.24 Non-EQ Electrical Cables and Connections Program NMP Unit 1 UFSAR Appendix C C-18 Rev. 25, October 2017 The Non-EQ Electrical Cables and Connections Program is a new program that manages aging of cables and connectors WSLR exposed to adverse localized temperature, moisture, or radiation environments. Program activities include periodic visual inspection of susceptible cables for evidence of cable and connection jacket surface anomalies. This program will be implemented prior to the period of extended operation. C.1.25 Non-EQ Electrical Cables and Connections Used in Instrumentation Circuits Program The Non-EQ Electrical Cables and Connections Used in Instrumentation Circuits Program manages aging of cables and connections exposed to adverse localized temperature and radiation environments that could result in loss of insulation resistance. It applies to accessible and inaccessible electrical cables that are not in the EQ Program and are used in circuits with sensitive, high-voltage, low-level signals such as radiation monitoring, nuclear instrumentation, and other such cables subject to aging management review that are sensitive to a reduction in insulation resistance. Activities include routine calibration tests of instrumentation loops, or direct testing of the cable system in those cases where cable testing is conducted as an alternate to surveillance testing, and in either case are implemented through the Surveillance Testing and Preventive Maintenance Programs. Testing is based on requirements of the particular calibrations, surveillances, or testing performed on the specific instrumentation circuit or cable and is implemented through the work control system. Where cable testing is conducted as an alternate to surveillance testing, the acceptance criteria for each test will be defined by the specific type of test performed and the specific cable tested. Enhancements to the Non-EQ Electrical Cables and Connections Used in Instrumentation Circuits Program include the following revisions to existing activities that are credited for license renewal: 1. Implement reviews of calibration or surveillance data for indications of aging degradation affecting instrument circuit performance. The first reviews will be completed prior to the period of extended operation and every 10 yr thereafter. 2. In cases where a calibration or surveillance program does not include the cabling system in the testing NMP Unit 1 UFSAR Appendix C C-19 Rev. 25, October 2017 circuit, or as an alternative to the review of calibration results described above, provide requirements and procedures to perform cable testing to detect deterioration of the insulation system, such as insulation resistance tests or other testing judged to be effective in determining cable insulation condition. The first test will be completed prior to the period of extended operation. The test frequency of these cables shall be determined based on engineering evaluation, but the test frequency shall be at least once every 10 yr. Enhancements will be completed prior to the period of extended operation. C.1.26 Non-Segregated Bus Inspection Program The Non-Segregated Bus Inspection Program manages aging effects for components and materials internal to the non-segregated bus ducts that connect the reserve auxiliary transformers to the 4160V buses required for the recovery of offsite power following a station blackout (SBO) event. Based upon the most recent industry and regulatory license renewal precedence, this program also includes normally energized bus ducts associated with boards feeding components WSLR. These normally-energized components are not subject to the EQ requirements of 10CFR50.49, but can be affected by elevated temperatures prior to the end of the period of extended operation. Program activities include 1) visual inspections of internal portions of the bus ducts to detect cracks, corrosion, debris, dust, and moisture; 2) visual inspections of the bus insulating system to detect embrittlement, cracking, melting, swelling, and discoloration; 3) visual inspections of bus supports (insulators) to detect cracking and lack of structural integrity; and 4) as an alternative to thermography or measuring connection resistance of bolted connections, a visual inspection for the accessible bolted connections that are covered with heat shrink tape, sleeving, insulating boots, etc. The program considers the technical information and guidance provided in applicable industry publications. Analytical trending is not included in this activity because the ability to trend inspection results is limited. This is an exception to the "Monitoring and Trending" element in Appendix A.1.2.3.5 to NUREG-1800. NMP Unit 1 UFSAR Appendix C C-20 Rev. 25, October 2017 Enhancements to the Non-Segregated Bus Inspection Program include expanded visual inspections of the bus ducts, their supports and insulation systems. Enhance program documents to develop acceptance criteria for visual inspection of the bus ducts, their supports and insulation systems, and the low-range ohmic checks of connections. Enhancements will be implemented prior to the period of extended operation. C.1.27 One-Time Inspection Program The One-Time Inspection Program is a new program that manages aging effects with potentially long incubation periods for susceptible components WSLR. Program activities include visual, volumetric, and other established inspection techniques consistent with industry practice to provide a means of verifying that an aging effect is either 1) not occurring, or 2) progressing so slowly that it has a negligible effect on the intended function of the structure or component. The program also provides measures for verifying the effectiveness of existing AMPs. This program is a new program that will be implemented prior to the period of extended operation. C.1.28 Open-Cycle Cooling Water System Program The Open-Cycle Cooling Water System (OCCWS) Program manages aging of components exposed to raw, untreated (e.g., service) water. For Unit 1, this includes portions of the service water system, the emergency service water system, shell side of the RBCLC heat exchangers, the EDG cooling water system, containment spray raw water system, and portions of the circulating water system. Also included are other components WSLR wetted by the service water system that are credited in the aging management review. The program also manages internal portions of non-safety related segments of the circulating water and service water systems which are WSLR per 10CFR54.4(a)(2). It also manages all aging effects for components subject to the scope of recommendations for GL 89-13. Program activities include 1) surveillance and control of biofouling (including biocide injection); 2) verification of heat transfer capabilities for components cooled by the service water system; 3) inspection and maintenance; 4) walkdown inspections; and 5) review of maintenance, operating, and NMP Unit 1 UFSAR Appendix C C-21 Rev. 25, October 2017 training practices and procedures. Inspections may include visual, UT, and eddy current testing (ECT) methods. This program is based on the recommendations of GL 89-13. Enhancements to the OCCWS Program include the following activities that are credited for license renewal: 1. Ensure that the applicable Unit 1 commitments made for GL 89-13, and the requirements in NUREG-1801, Section XI.M20, are captured in the Unit 1 implementing documents for GL 89-13. 2. Where the requirements of the NUREG-1801, Section XI.M20, are more conservative than the GL 89-13 commitments, they will be incorporated into the OCCWS Program. 3. Revise Unit 1 preventive maintenance and heat transfer performance test procedures to incorporate specific inspection criteria, corrective actions, and frequencies. Enhancements will be completed prior to the period of extended operation. C.1.29 Preventive Maintenance Program The scope of the Preventive Maintenance (PM) Program includes, but is not limited to, valve bodies, heat exchangers, expansion joints, tanks, ductwork, fan/blower housings, dampers, and pump casings. This program provides for performance of various maintenance activities on a specified frequency based on vendor recommendations and operating experience. These activities provide opportunities for component condition monitoring to manage the effects of aging for many SSCs WSLR. Enhancements to the PM Program include the following revisions to existing activities that are credited for license renewal: 1. Expand the PM Program to encompass activities for certain additional components identified as requiring aging management. 2. Explicitly define the aging management attributes, including the systems and the component types/commodities included in the program. NMP Unit 1 UFSAR Appendix C C-22 Rev. 25, October 2017 3. Specifically list activities credited for aging management, parameters monitored, and the aging effects detected. 4. Establish a requirement that inspection data be monitored and trended. 5. Establish detailed parameter-specific acceptance criteria. Enhancements will be completed prior to the period of extended operation. C.1.30 Reactor Head Closure Studs Program The Reactor Head Closure Studs Program manages cracking and loss of material from the RPV closure studs. This program implements the preventive measures of RG 1.65. Inservice examinations are performed in accordance with the 1989 edition of the ASME Boiler and Pressure Vessel Code with no addenda, and ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," 1995 edition through 1996 addenda as approved by the NRC in plant-specific exemptions. This is an exception to the program described in NUREG-1801 (which cites ASME Section XI requirements covered in the 1995 edition through 1996 addenda). C.1.31 Reactor Vessel Surveillance Program The Reactor Vessel Surveillance Program manages loss of fracture toughness due to neutron irradiation embrittlement in the RPV beltline material. Program activities include 1) periodic withdrawal and testing of surveillance capsules from the RPV; 2) use of test results and allowable stress loadings for the ferritic RPV materials to determine operating limits; and 3) comparison with a large industry data set to confirm validity of test results. Analysis and testing are based on the requirements of 10CFR50 Appendix H, and ASTM Standard E-185. Nine Mile Point commits to implement the Integrated Surveillance Program (ISP) described in BWRVIP-116 (if approved by the NRC staff). When the NRC issues a final safety evaluation report (SER) for BWRVIP-116, Nine Mile Point will address any open items and complete the SER action items. Should BWRVIP-116 not be approved by the NRC, a plant-specific Reactor Vessel Surveillance Program will be submitted to the NRC 2 yr prior to commencement of the period of extended operation. NMP Unit 1 UFSAR Appendix C C-23 Rev. 25, October 2017 Enhancements to the Reactor Vessel Surveillance Program include the following revisions to existing activities that are credited for license renewal: 1. Incorporate the requirements and elements of the ISP, as documented in BWRVIP-116, if approved by the NRC or an NRC-approved plant-specific program, into the Reactor Vessel Surveillance Program, and include a requirement that if Unit 1 surveillance capsules are tested, the tested specimens will be stored in lieu of optional disposal. 2. Project analyses of upper-shelf energy (USE) and pressure-temperature (P-T) limits to 60 yr using methods prescribed by RG 1.99, Revision 2, and include the applicable bounds of the data, such as operating temperature and neutron fluence. Enhancements will be completed prior to the period of extended operation. C.1.32 Selective Leaching of Materials Program The Selective Leaching of Materials Program is a new program that manages aging of components susceptible to selective leaching. The potentially susceptible components include valve bodies, valve bonnets, pump casings, and heat exchanger components in various systems. This program will be implemented through the One-Time Inspection Program prior to the period of extended operation. C.1.33 Structures Monitoring Program The Structures Monitoring Program manages aging of structures, structural components, and structural supports WSLR. The program provides for periodic visual inspections, surveys, and examination of all safety-related buildings (including the primary containment and substructures within the primary containment), and various other buildings WSLR. Program activities identify degradation of materials of construction, which include structural steel, concrete, masonry block, sealing materials, and a Unit 1 wooden structure. While not credited for mitigation of aging, protective coatings are also inspected under this program. The Structures Monitoring Program, which was initially developed to meet the regulatory requirements of NMP Unit 1 UFSAR Appendix C C-24 Rev. 25, October 2017 10CFR50.65, implements guidance provided in RG 1.160, NUMARC 93-01, and NEI 96-03. Enhancements to the Structures Monitoring Program include the following revisions to existing activities that are credited for license renewal: 1. Expand the parameters monitored during structural inspections to include those relevant to aging effects requiring management identified for structural bolting. 2. Implement regularly scheduled groundwater monitoring to ensure that a benign environment is maintained. 3. Expand the scope of the program to include the steel electrical transmission towers required for the SBO recovery path that are WSLR, but not within the current scope of 10CFR50.65. 4. The Masonry Wall Program (as managed by the Structures Monitoring Program) will be enhanced to provide guidance for inspecting non-reinforced masonry walls that do not have bracing and are WSLR more frequently than the reinforced masonry walls. Enhancements will be completed prior to the period of extended operation. C.1.34 Systems Walkdown Program The Systems Walkdown Program manages aging effects for accessible external surfaces of pumps, valves, piping, bolts, heat exchangers, tanks, heating, ventilation and air conditioning (HVAC) components, and other components. Visual inspections identify corrosion, changes in material properties, signs of material degradation, and leakage. The program also identifies adverse conditions that can lead to aggressive environments for systems and components within the scope of license renewal. Program activities include system engineer walkdowns (i.e., field evaluations of system components to assess material condition), documentation and evaluation of inspection results, and appropriate corrective actions. Enhancements to the Systems Walkdown Program include the following revisions to existing activities that are credited for license renewal: NMP Unit 1 UFSAR Appendix C C-25 Rev. 25, October 2017 1. Train all personnel performing inspections in the Systems Walkdown Program to ensure that age-related degradation is properly identified, and incorporate this training into the site training program. 2. Specify acceptance criteria for visual inspections to ensure aging-related degradation is properly identified and corrected. Enhancements will be completed prior to the period of extended operation. C.1.35 Torus Corrosion Monitoring Program The Torus Corrosion Monitoring Program manages corrosion of the Unit 1 suppression chamber (torus) through inspection and analysis. This program provides for 1) determination of torus shell thickness through ultrasonic measurement; 2) determination of corrosion rate through analysis of material coupons; and 3) visual inspection of accessible external surfaces of the torus support structure for corrosion. The Torus Corrosion Monitoring Program ensures that the Unit 1 torus shell and support structure thickness limits are not exceeded. C.1.36 Water Chemistry Control Program The Water Chemistry Control Program manages aging effects by controlling the internal environment of the reactor water, feedwater, condensate, and control rod drive systems, and related auxiliaries (such as the torus, condensate storage tank, and spent fuel pool). The aging effects of concern are 1) loss of material; and 2) crack initiation and growth. Program activities include monitoring and controlling concentrations of known detrimental chemical species below the levels known to cause degradation. The Water Chemistry Control Program implements the guidelines for BWR water chemistry presented in EPRI Reports TR-103515-R1 and TR-103515-R2. This is an exception to the program described in NUREG-1801 (which identifies EPRI TR-103515-R0 as the basis for BWR water chemistry programs). C.1.37 Bolting Integrity Program The Bolting Integrity Program manages aging effects due to loss of preload, cracking, and loss of material of bolting within the scope of license renewal, including safety-related bolting, NMP Unit 1 UFSAR Appendix C C-26 Rev. 25, October 2017 bolting for nuclear steam supply system (NSSS) component supports, bolting for other pressure-retaining components, and structural bolting. Program activities include periodic inspections of bolting for indication of loss of preload, cracking, and loss of material due to corrosion, rust, etc. This program is based on the guidelines delineated in NUREG-1339 and the guidance contained in EPRI NP-5769, with exceptions noted in NUREG-1339, for safety-related bolting and EPRI TR-104213 for other bolting. The Bolting Integrity Program is implemented through the ASME Section XI Inservice Inspection (Subsections IWB, IWC, IWD) Program, ASME Section XI Inservice Inspection (Subsection IWE) Program, ASME Section XI Inservice Inspection (Subsection IWF) Program, Structures Monitoring Program, Preventive Maintenance Program, and Systems Walkdown Program. An exception is taken to the GALL report in that Unit 1 utilizes the 1989 edition with no addenda of the ASME Section XI Code versus the 1995 edition through the 1996 addenda. Enhancements to the Bolting Integrity Program include the following: 1. Establish an augmented inspection program for high-strength (actual yield strength 150 ksi) bolts. This augmented program will prescribe the examination requirements of Tables IWB-2500-1 and IWC-2500-1 of ASME Section XI for high-strength bolts in the class 1 and class 2 component supports, respectively. 2. The Structures Monitoring, Preventive Maintenance, and Systems Walkdown Programs will be enhanced to include requirements to inspect bolting for indication of loss of preload, cracking, and loss of material, as applicable. 3. Include in administrative and implementing program documents references to the Bolting Integrity Program and industry guidance. Enhancements will be completed prior to the period of extended operation. C.1.38 BWR Control Rod Drive Return Line Nozzle Program The Unit 1 BWR Control Rod Drive Return Line (CRDRL) Nozzle Program is an existing program that requires UT inspections of the CRDRL nozzle every 10 yr to verify the nozzle is acceptable for continued service. A CRDRL crack growth fracture mechanics NMP Unit 1 UFSAR Appendix C C-27 Rev. 25, October 2017 analysis was used to demonstrate the adequacy of the 10-yr inspection frequency. The crack growth analyses are TLAAs that are managed in accordance with 10CFR54.21(c)(1)(iii), as described in Section 4.3.3. The three exceptions to NUREG-1801, Section XI.M6, are: 1. The Unit 1 Inservice Inspection (ISI) Program does not comply with the specific edition and addenda of ASME Section XI cited in the GALL report because the program is updated to the latest edition and addenda of ASME Section XI, as mandated by 10CFR50.55a, prior to the start of each inspection interval; 2. The Unit 1 program uses enhanced ultrasonic inspection techniques instead of PT inspections to satisfy the recommendations of NUREG-0619 (now superseded by Appendix VIII to ASME Section XI, Division 1, 1995 edition with the 1996 addenda); and, 3. The Unit 1 program uses an inspection frequency of every 10 yr versus every sixth refueling outage or 90 startup/shutdown cycles specified in NUREG-0619. C.1.39 Protective Coating Monitoring and Maintenance Program The Protective Coating Monitoring and Maintenance Program is described in the Unit 1 response to GL 98-04, "Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-Of-Coolant Accident because of Construction and Protective Coating Deficiencies and Foreign Material in Containment." The program applies to Service Level 1 protective coatings inside the primary containment and items within the torus (outside surface of the vent [ring] header and downcomer, inside surface of the vent piping, ring header, vent header junctions, and downcomers). The condition assessments and resulting repair, replacement, or removal activities ensure that the amount of coatings subject to detachment from the substrate during a loss-of-coolant accident (LOCA) is minimized to ensure post-accident operability of the emergency core cooling system (ECCS) suction strainers. The Protective Coating Monitoring and Maintenance Program takes exception to certain NUREG-1801, Section XI.S8 (Protective Coating Monitoring and Maintenance Program) evaluation elements, in that it is not credited for prevention of corrosion of carbon steel. The program will be enhanced following the guidance NMP Unit 1 UFSAR Appendix C C-28 Rev. 25, October 2017 within ASTM D5163-05a, and measurements of cracks, peeling, or delaminated coatings will be estimated via visual methods. Planned program enhancements include the following: 1. Specifying the visual examination of coated surfaces for any visible defects including blistering, cracking, flaking, peeling, and physical or mechanical damage. 2. Performance of periodic inspection of coatings in the Drywell every refueling outage versus every 24 months. Inspection of coatings in the Torus will be performed every other refueling outage. 3. Setting minimum qualifications for inspection personnel, the inspection coordinator, and the inspection results evaluator. 4. Performing the thorough visual inspection and areas noted as deficient along with the general visual inspection. 5. Specifying the types of instruments and equipment that may be used for the inspection. 6. Requiring pre-inspection reviews of the previous two monitoring reports before performing the condition assessment. 7. Establishing guidelines for prioritization of repair areas and monitoring these areas until they are repaired. 8. Requiring that the inspection results evaluator determine which areas are not acceptable and initiates corrective action. Enhancements will be completed prior to the period of extended operation. C.1.40 Non-EQ Electrical Cable Metallic Connections Inspection Program The Non-EQ Electrical Cable Metallic Connections Inspection Program is a new plant-specific program that manages the aging effects of the metallic portion of electrical cable connections NMP Unit 1 UFSAR Appendix C C-29 Rev. 25, October 2017 that are not subject to the qualification requirements of 10CFR50.49, but are still subject to aging effects caused by various stressors. These aging stressors include: thermal cycling, ohmic heating, electrical transients, vibration, chemical contamination, corrosion, and oxidation. All connections associated with cables that are in scope for license renewal are part of this program. This program is a one-time inspection program, on a representative sample basis, of the non-EQ electrical cable metallic connection population subject to aging management to ensure that aging that would affect the ability of the non-EQ electrical cable metallic connections to perform their intended function is not occurring. The one-time inspections will be completed at least once prior to the period of extended operation to verify that loosening and/or high resistance of the non-EQ electrical cable connections due to the identified potential aging stressors is not occurring and, therefore, periodic inspections are not required during the period of extended operation. Trending actions are not included as part of this program because it is a one-time inspection program. C.1.41 Drywell Supplemental Inspection Program The Drywell Supplemental Inspection Program manages the aging effects of localized areas of the Unit 1 drywell shell identified as having major corrosion in the Unit 1 Owner Activity Report dated July 23, 2003. Volumetric examinations will be performed during the 2007 refueling outage, and an engineering evaluation will be performed to determine what actions, beyond those required by the ASME Section XI Inservice Inspection (Subsection IWE) Program, are necessary for operation through the period of extended operation. Corrective actions could include increased monitoring, application of a protective coating, repair or replacement of affected sections, or other actions deemed appropriate by Engineering. The Unit 1 Drywell Supplemental Inspection Program is a new program that will be implemented prior to the period of extended operation. C.2 TIME-LIMITED AGING ANALYSES SUMMARIES As part of the application for a renewed license, 10CFR54.21(c) requires that an evaluation of TLAAs for the period of extended operation be provided. The following TLAAs have been identified and evaluated to meet this requirement. NMP Unit 1 UFSAR Appendix C C-30 Rev. 25, October 2017 C.2.1 Reactor Vessel Neutron Embrittlement Analysis The ferritic materials of the reactor vessel are subject to embrittlement due to high-energy neutron exposure. The evaluation of reactor vessel neutron embrittlement is a TLAA. The following TLAA discussions are related to the issue of neutron embrittlement.

• Upper-shelf energy
• Pressure-temperature limits
• Elimination of circumferential weld inspection
• Axial weld failure probability C.2.1.1 Upper-Shelf Energy Ferritic RPV materials undergo a transition in fracture behavior from brittle to ductile as the temperature of the material is increased. Charpy V-notch tests are conducted in the nuclear industry to monitor changes in the fracture behavior during irradiation. Neutron irradiation to fluences above approximately 1x1017 n/cm2 causes an upward shift in the ductile-to-brittle transition temperature and a drop in USE. To satisfy the acceptance criteria for USE contained in 10CFR50 Appendix G, the RPV beltline materials must have a Charpy USE of no less than 50 ft-lbs throughout the life of the RPV unless it can be demonstrated that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code. The USE for the limiting beltline weld materials for Unit 1 is predicted to remain above 50 ft-lbs throughout the period of extended operation based on projected fluence values. The USE of the limiting plate material for Unit 1 is below 50 ft-lbs but is predicted to remain above the value required by an equivalent margins analysis based on projected fluence values. Therefore, the USE for the Unit 1 RPV beltline materials has been projected (reevaluated) for the period of extended operation in accordance with 10CFR54.21(c)(1)(ii). C.2.1.2 Pressure-Temperature Limits 10CFR50 Appendix G requires that the RPV be operated within established P-T limits during heatup and cooldown. These limits NMP Unit 1 UFSAR Appendix C C-31 Rev. 25, October 2017 specify the maximum allowable pressure as a function of reactor coolant temperature. Unit 1 Technical Specifications did contain P-T limit curves for heatup, cooldown, inservice leakage testing, and hydrostatic testing, and limit the maximum rate of change of reactor coolant temperature at the time the License Renewal Application was approved. Subsequently, the P-T limit curves were relocated to a Pressure and Temperature Limits Report (PTLR). The P-T limit curves are periodically revised to account for changes in fracture toughness of the RPV components due to anticipated neutron embrittlement effects for higher accumulated fluences. Calculation of P-T limit curves using the projected fluence at the end of the period of extended operation would result in unnecessarily restrictive operating curves. However, projection of the adjusted reference temperature (ART), which is used in development of the curves, to the end of the period of extended operation provides assurance that development of P-T limit curves will be feasible up to the maximum predicted effective full power year (EFPY). Projections of the ART values for the beltline materials have been made for the period of extended operation, providing reasonable assurance that it will be possible to prepare P-T curves that will permit continued plant operation. The P-T curves (and the related PTLR) will continue to be updated, either as required by 10CFR50 Appendix G to assure the operational limits remain valid at the current cumulative neutron fluence levels, or on an as-needed basis to provide appropriate operational flexibility. C.2.1.3 Elimination of Circumferential Weld Inspection Relief from reactor vessel circumferential weld examination requirements under GL 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," is based on probabilistic assessments that predict an acceptable probability of failure per reactor operating year. The analysis is based on reactor vessel metallurgical conditions as well as flaw indication sizes and frequencies of occurrence that are expected at the end of a licensed operating period. Unit 1 has received relief from reactor vessel circumferential weld examination requirements under GL 98-05 for the remainder of its current 40-yr license term (Reference 1).

NMP Unit 1 UFSAR Appendix C C-32 Rev. 25, October 2017 Projected values of mean and upper bound reference temperature nil ductility transition temperature (RTNDT) for the limiting circumferential welds at Unit 1 are below the bounding mean RTNDT determined by the NRC staff in the SER for BWRVIP-05 (Reference 7). Thus, there is reasonable assurance the conditional probability of vessel failure due to Unit 1 RPV circumferential weld failure is bounded by the NRC analysis. Unit 1 will apply for relief from circumferential weld inspections for the period of extended operation. Supporting analyses, procedural controls, and operator training will be completed prior to the period of extended operation to support and confirm that the RPV circumferential weld failure probability remains acceptable for the period of extended operation. Based on the scoping evaluation discussed above, there is reasonable assurance the failure probability will remain acceptable for the period of extended operation. C.2.1.4 Axial Weld Failure Probability In the safety evaluation presented in "Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report" (Reference 8), the NRC staff indicates that the RPV failure frequency due to failure of the limiting axial welds in the BWR fleet at the end of 40 yr of operation is less than 5x10-6 per reactor year, given the assumptions on flaw density, distribution, and location described in the SER. Projected values of mean RTNDT and upper bound RTNDT for the limiting axial welds at Unit 1 are below the bounding mean RTNDT value determined by the NRC staff in the SER for BWRVIP-74-A (Reference 2). Thus, there is reasonable assurance that the RPV failure frequency due to failure of the limiting axial weld is expected to remain less than 5x10-6 per reactor year for Unit 1 during the period of extended operation. Inspection of the axial welds in accordance with the ASME XI Code requirements will continue at Unit 1 during the period of extended operation. Supporting analyses will be completed prior to the period of extended operation to confirm that the RPV axial weld failure probability for the limiting Unit 1 axial weld remains bounded for the period of extended operation. Based on the scoping evaluation discussed above, there is reasonable assurance the failure probability will remain acceptable for the period of extended operation. C.2.2 Metal Fatigue Analysis NMP Unit 1 UFSAR Appendix C C-33 Rev. 25, October 2017 ASME Section III requires calculation of CUFs to demonstrate fatigue-tolerant design for reactor vessels, vessel internals, Class 1 piping and components, metal containments, and penetrations. These values are indexed to the number of transients anticipated over the design life of the component (usually 40 yr). Designated plant events have been counted and categorized to ensure that the number of actual operational transient cycles does not exceed the number of transients assumed in the plant design for fatigue. For certain events that affect fatigue usage, linear projections of the actual data to the end of the period of extended operation will exceed the analyzed number of design basis transients. For those locations where additional fatigue analysis is required to take advantage of the implicit margin (and to more accurately determine CUFs), the EPRI FatiguePro fatigue monitoring software will be implemented. The following thermal and mechanical fatigue analyses of mechanical components have been identified as TLAAs: 1. Reactor Vessel Fatigue Analysis 2. Feedwater Nozzle and Control Rod Drive Return Line Nozzle Fatigue and Cracking Analyses 3. Non-ASME Section III Class 1 Piping and Components Fatigue Analysis 4. Reactor Vessel Internals Fatigue Analysis 5. Environmentally Assisted Fatigue 6. Fatigue of the Emergency Condenser C.2.2.1 Reactor Vessel Fatigue Analysis The original design of RPV pressure boundary components included analyses of fatigue resistance. Components were evaluated by calculating the alternating stresses associated with applicable design transients and determining a CUF based on the number of anticipated transients for the original 40-yr life of the plant. Fatigue-tolerant design is demonstrated for those locations with CUFs less than 1.0. For the critical RPV component locations, transients contributing to fatigue usage will be tracked by the FMP NMP Unit 1 UFSAR Appendix C C-34 Rev. 25, October 2017 (Section C.1.16) with additional usage added to the baseline CUF. The FMP provides an analytical basis for confirming that the number of cycles established by the analysis of record will not be exceeded before the end of the period of extended operation. C.2.2.2 Feedwater Nozzle and Control Rod Drive Return Line Nozzle Fatigue and Cracking Analyses BWRs have experienced fatigue crack initiation and growth in feedwater system and CRDRL nozzles. Rapid thermal cycling (occurring as a result of bypass leakage past loose-fitting thermal sleeves, or in nozzles lacking thermal sleeves) initiated fatigue cracks that propagated due to larger (in terms of the magnitude of temperature and pressure change) thermal cycles resulting from plant transients. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," identifies interim and long-term procedural and design changes to minimize thermal fatigue cracking, as well as inspection requirements. Various calculations were prepared in response to NUREG-0619 (e.g., to support enhanced inspection intervals, to incorporate updated fatigue crack growth curves, etc.), and CUFs were determined on the basis of anticipated transients for the original 40-yr life of the plant. Fatigue-tolerant design is demonstrated for those locations with CUFs less than 1.0. The Unit 1 feedwater nozzles require continued monitoring (including analysis using FatiguePro) to demonstrate compliance over the period of extended operation. Transients contributing to fatigue usage of the feedwater nozzles will be tracked by the FMP (Section C.1.16) with additional usage added to the baseline CUF. Additionally, the Unit 1 feedwater nozzles will be periodically inspected in accordance with Unit 1 commitments related to NUREG-0619. The fatigue usage of the Unit 1 CRDRL nozzle has been calculated to be significantly below the allowable fatigue usage of 1.0 over the life of the plant, including a 20-yr license extension. However, Unit 1 will continue to perform enhanced inspections of the CRDRL nozzle in accordance with commitments to NUREG-0619. C.2.2.3 Non-ASME Section III Class 1 Piping and Components Fatigue Analysis Piping and components WSLR were designed to codes other than ASME Section III Class 1. Applicable codes include ASA NMP Unit 1 UFSAR Appendix C C-35 Rev. 25, October 2017 B31.1-1955. These codes do not require explicit fatigue analyses. Instead, the effects of cyclic loading are accounted for through application of stress range reduction factors based on the anticipated number of equivalent full temperature thermal expansion cycles over the original 40-yr life of the plant. The original design for cyclic loading is expected to remain valid for the period of extended operation for the majority of non-ASME Class 1 systems and components. However, non-ASME Class 1 locations meeting one or more of the following criteria require development of fatigue analyses (similar to those performed for ASME Class 1 piping): 1. The location experiences high fatigue usage due to significant thermal transients due primarily to on/off flow, stratification, and local thermal cycling effects; 2. The location experiences high fatigue usage due to structural or material discontinuities that result in high stress indices (e.g., at thickness transitions); 3. The location has been identified in NUREG/CR-6260 (Reference 3) for the older-vintage BWRs (i.e., locations equivalent to the recirculation line at the RHR return line tee, the RHR line at the tapered transition, and the feedwater line at the RCIC tee). Based on the above criteria, portions of the following Unit 1 systems were identified for further analysis: 1. Feedwater/high-pressure coolant injection (HPCI) system; 2. Core spray system; 3. RWCU system (piping inside the RCPB); and 4. Reactor recirculation system (and associated shutdown cooling system lines). Prior to the period of extended operation, a baseline CUF (based on a conservative analysis of the fatigue usage to-date) will be determined for the specified portions of the Unit 1 systems listed above. If the baseline CUF for a specified portion of a system exceeds 0.4 (considered a general threshold of significance), the limiting location may require monitoring to NMP Unit 1 UFSAR Appendix C C-36 Rev. 25, October 2017 demonstrate compliance over the period of extended operation. For the limiting locations, those transients contributing to fatigue usage will be tracked by the FMP with additional usage added to the baseline CUF. C.2.2.4 Reactor Vessel Internals Fatigue Analysis Determination of CUFs was not a design requirement for reactor vessel internals at Unit 1. However, core shroud stabilizer assemblies (tie-rods) and mechanical clamps installed as repairs for cracked horizontal and vertical core shroud welds were evaluated for fatigue using ASME Section III methods to calculate alternating stresses and determine CUF values. Fatigue-tolerant design is demonstrated for the tie-rods and mechanical clamps with CUFs less than 1.0. The potential for cracking of components comprising the reactor vessel internals, both due to fatigue and (more significantly) IGSCC, is managed by the BWRVIP (Section C.1.12), which incorporates comprehensive inspection and evaluation guidelines issued by the BWRVIP and approved by the NRC. These activities provide assurance that any unexpected degradation resulting from fatigue in the reactor vessel internals for the current license period and the period of extended operation will be identified and corrected. Therefore, the effects of fatigue on the intended function(s) of the reactor vessel internals will be adequately managed for the period of extended operation. C.2.2.5 Environmentally-Assisted Fatigue Generic Safety Issue (GSI) 190, "Fatigue Evaluation of Metal Components for 60-Year Plant Life," was established to address NRC concerns regarding environmental effects on fatigue of pressure boundary components for 60 yr of plant operation. The NRC staff studied the probability of fatigue failure for selected metal components based on the increased CUFs determined in NUREG/CR-6260 (Reference 3) and a 60-yr plant life. The NRC closed this GSI and concluded that environmental effects did not substantially affect core damage frequency. However, since the nature of age-related degradation indicated the potential for an increase in the frequency of pipe leaks as plants continue to operate, licensees are required to address the effects of coolant environment on component fatigue life as AMPs are formulated in support of license renewal. Unit 1 will assess the impact of the reactor coolant environment on a sample of critical component locations, including locations NMP Unit 1 UFSAR Appendix C C-37 Rev. 25, October 2017 equivalent to those identified in NUREG/CR-6260 as part of the FMP (Section C.1.16). These locations will be evaluated by applying environmental correction factors (Fen) to existing and future fatigue analyses. Evaluation of the sample of critical components will be completed prior to the period of extended operation. C.2.2.6 Fatigue of the Emergency Condenser The emergency cooling system provides for decay heat removal from the reactor fuel in the event that reactor feedwater capability is lost and the main condenser is unavailable. The tube and shell sides of the emergency condensers were designed in accordance with ASME Section III Class 2 and 3, respectively. The original tubing has experienced thermal fatigue resulting from leakage past the condensate return valve to the RPV. As part of the subsequent modification and repair, fatigue loading was evaluated by calculating the alternating stresses associated with applicable design transients and determining a CUF based on the number of anticipated transients for the life of the condensers. Fatigue-tolerant design is demonstrated for components with CUFs less than 1.0. While the CUFs were shown to be less than 1.0, certain locations in the Unit 1 emergency condensers require continued monitoring (including analysis using FatiguePro) to demonstrate compliance over the period of extended operation. The FMP (Section C.1.16) will track transients specific to the emergency cooling system with additional usage added to the baseline CUF for the condensers. C.2.3 Environmental Qualification The following EQ analysis has been identified as a TLAA: Electrical Equipment EQ C.2.3.1 Electrical Equipment EQ 10CFR50.49 requires that certain safety-related and non-safety related electrical equipment remain functional during and after identified design basis events. To establish reasonable assurance that this equipment can function when exposed to postulated harsh environmental conditions, licensees are required to determine the equipment's qualified life and to develop a program that maintains the qualification of that equipment. NMP Unit 1 UFSAR Appendix C C-38 Rev. 25, October 2017 For components within the scope of the EQ Program (Section C.1.15), analyses of thermal exposure, radiation exposure, and mechanical cycle aging that cannot be shown to remain valid for the period of extended operation will be projected to extend the qualification of components before reaching the aging limits established in the applicable evaluation, or the components will be refurbished or replaced. C.2.4 Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis The following containment liner plate, metal containments, and penetrations fatigue analyses have been identified as TLAAs:

• Torus Shell and Vent System Fatigue Analysis
• Torus-Attached Piping Analysis
• Torus Wall Thickness
• Fatigue of Primary Containment Penetrations C.2.4.1 Torus Shell and Vent System Fatigue Analysis Large-scale testing of the Mark III containment and in-plant testing of Mark I primary containment systems identified additional hydrodynamic loads that were not considered in the original design of the Mark I containment used at Unit 1. To provide the bases for generic load definition and structural assessment techniques, General Electric Company (GE) initiated the Mark I Containment Program. NUREG-0661, "Safety Evaluation Report, Mark I Containment Long Term Program, Resolution of Generic Technical Activity A-7," requires a plant-unique analysis for each Mark I configuration to evaluate the effects of the hydrodynamic stresses resulting from a LOCA and SRV discharge. The 60-yr CUF values for the controlling locations in the torus shell are less than 1.0. Therefore, the Unit 1 torus shell has been evaluated and is qualified for the period of extended operation. C.2.4.2 Torus-Attached Piping Analysis As a result of the Mark I Containment Program, modifications were performed at Unit 1, including changes to the configuration NMP Unit 1 UFSAR Appendix C C-39 Rev. 25, October 2017 of SRV piping and other piping penetrating the suppression chamber (torus) (generically referred to herein as torus-attached piping). As part of the generic Mark I Containment Program, fatigue analyses were performed considering the design loads identified in NUREG-0661 and its supplements. Fatigue-tolerant design is demonstrated for those locations with CUFs less than 1.0. The bounding 40-yr CUFs for the subject piping and associated penetrations are less than 0.5; therefore, the 60-yr CUF values for all controlling locations can be demonstrated to remain less than 1.0. However, SRV actuations, which are the only non-accident or earthquake contributor to torus-attached piping fatigue usage, have not been counted historically. SRV actuations for Unit 1 to date have been estimated. To ensure that the fatigue usage of the torus-attached piping remains within design values, SRV actuations will be added to the FMP (Section C.1.16) as a transient that is monitored. The two torus-attached piping locations with the highest calculated fatigue usage will be added to the FMP as locations to be monitored. Therefore, the effects of fatigue on the Unit 1 torus-attached piping will be adequately managed for the period of extended operation. C.2.4.3 Torus Wall Thickness The Unit 1 suppression chamber (torus) is constructed of A201 Grade B (firebox) steel plates with a certified minimum thickness of 0.460 in. This value included an original corrosion allowance of 0.0625 in, which was added to the minimum wall thickness required by the applicable design codes. However, subsequent addition of hydrodynamic loads (resulting from LOCA and SRV actuation) to the containment design bases resulted in a reduction of the corrosion allowance. To establish reasonable assurance that the revised minimum wall thickness of 0.431 in is not reached, Unit 1 is required to monitor torus wall thickness and corrosion rate (Reference 4). Determination of torus corrosion rates is an ongoing activity that considers inspection results and the remaining corrosion allowance. The Torus Corrosion Monitoring Program (Section C.1.35) has been developed to monitor the torus shell material thickness and ensure it is maintained within the bounds of the qualification bases. Therefore, the effects of loss of material on the intended function(s) of the torus shell will be adequately managed during the period of extended operation.

NMP Unit 1 UFSAR Appendix C C-40 Rev. 25, October 2017 C.2.4.4 Fatigue of Primary Containment Penetrations The Unit 1 drywell was designed as a Class B vessel in accordance with Section III of the ASME Boiler and Pressure Vessel (B&PV) Code, 1965 edition (ASME Section III, 1965). The 1965 edition of the ASME Section III B&PV Code did not require fatigue analysis of Class B vessels. The drywell penetrations were considered an extension of the drywell and thus did not require fatigue analysis. For Unit 1, fatigue of torus penetrations was addressed in the same analysis as the torus-attached piping, the "Plant Unique Analysis Report of the Torus Attached Piping for Nine Mile Point Unit 1 Nuclear Generation Station," which was transmitted to the NRC in a letter dated May 22, 1984. This analysis was performed in accordance with ASME Section III, 1977 edition through the summer 1977 addenda. Fatigue analyses were performed for the SRV penetration (where the SRV line penetrates the vent header spherical intersection) and torus-attached piping penetrations. The fatigue analyses for the SRV and torus-attached piping penetrations considered a number of cycles related to anticipated transients for the original 40-yr life of the plant. The number of anticipated significant transient cycles for a 40-yr life divided by the maximum number of allowable cycles for the transient producing the maximum stress was used to estimate the 40-yr design CUF. Linear projection of this CUF to 60 yr results in a CUF far below the allowable. C.2.5 Other Plant-Specific TLAAs The following plant-specific TLAAs have been identified for Unit 1:

• Reactor Vessel and Reactor Vessel Closure Head Weld Flaw Evaluations
• RWCU System Weld Overlay Fatigue Flaw Growth Evaluations C.2.5.1 Reactor Vessel and Reactor Vessel Closure Head Weld Flaw Evaluations During refueling outage (RFO) 15, augmented examinations identified unacceptable flaw indications in two RPV shell welds (Reference 5). During RFO17, UT examinations identified an NMP Unit 1 UFSAR Appendix C C-41 Rev. 25, October 2017 unacceptable flaw indication in a closure head meridional weld (Reference 6). Structural evaluations of these flaws (performed in accordance with ASME Section XI, Subsection IWB-3600) compared the flaw characteristics to predetermined acceptability criteria to justify continued operation without repair of the flaw. Since the acceptability criteria were based on an assumed number of transient cycles applicable to the original 40-yr license term, the subject evaluations satisfy the criteria of 10CFR54.3(a). The number of cycles from the time of inspection to the end of the evaluation period is used to determine crack growth. With the addition of the period of extended operation (20 yr), the Unit 1 RPV can be expected to accumulate fatigue usage for no more than an additional 25 yr. During this interval, it is unlikely that the number of startup/shutdown cycles that occur will result in exceeding the 240 additional startup/shutdown cycles that were the bases for the evaluation. The actual interval is the period of time from the date of the inspection (March 2003) through the end of the period of extended operation. Therefore, the RPV closure head weld flaw evaluation remains valid for the period of extended operation. No later than 2 yr prior to the period of extended operation, the RPV weld flaw evaluation will be revised to consider additional fatigue crack growth and the effects of additional irradiation embrittlement associated with operation for an additional 20 yr, and submitted to the NRC for review and approval. The flaws will be reexamined in accordance with ASME Section XI as necessary. C.2.5.2 Reactor Water Cleanup System Weld Overlay Fatigue Flaw Growth Evaluations Fatigue crack growth analyses have been performed for two weld overlays in the RWCU system. The repaired welds are located at the inlet nozzle of the regenerative heat exchanger and the transition pipe between the upper and lower shells of the regenerative heat exchanger, respectively. The weld overlays consist of IGSCC-resistant austenitic stainless steel material and, thus, are not susceptible to continued IGSCC crack propagation. However, the first 1/16-in thick layer of weld metal deposited is not assumed to be IGSCC-resistant due to weld dilution; thus, it is assumed to be cracked. A fatigue crack growth analysis for each weld overlay was performed in accordance with ASME Section XI, Appendix C, with the crack propagating into the overlay from the hypothetical 1/16-in deep crack. The results of those analyses showed that the welds were acceptable per the Code criteria through the end of the period NMP Unit 1 UFSAR Appendix C C-42 Rev. 25, October 2017 of extended operation. Additionally, however, the overlaid welds are UT examined periodically under the BWR Stress Corrosion Cracking Program, thus ensuring there is no fatigue crack propagation into the overlays. The maximum interval between inspections is defined by the requirements of BWRVIP-75-A. Therefore, the aging of the RWCU weld overlays will be adequately managed through the balance of the initial 40-yr licensing term and the period of extended operation. C.3 GENERIC QUALITY ASSURANCE PROGRAM REQUIREMENTS FOR LICENSE RENEWAL The Nine Mile Point Quality Assurance Program implements the requirements of 10CFR50 Appendix B, and is consistent with the summary in Appendix A.2 of NUREG-1800, "Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants," published July 2001. The elements of corrective action, confirmation process, and administrative controls in the Quality Assurance Program are applicable to both safety-related and non-safety related SSCs that are subject to an aging management review. Generically, these three elements are applicable as follows: 1. Corrective Actions Corrective actions are implemented in accordance with the requirements of 10CFR50 Appendix B, as committed to in the Quality Assurance Topical Report (QATR). The Corrective Action Program provides for the identification, evaluation, and resolution of nonconforming conditions. 2. Confirmation Process The confirmation process is part of the Corrective Action Program, which is implemented in accordance with the requirements of 10CFR50 Appendix B, as committed to in the QATR. The focus of the confirmation process is on the verification that corrective actions are effective. The measure of effectiveness is in terms of correcting the adverse condition and precluding repetition of significant conditions adverse to quality. 3. Administrative Controls NMP Unit 1 UFSAR Appendix C C-43 Rev. 25, October 2017 AMPs are implemented through various plant documents. These implementing documents are subject to administrative controls, including a formal review and approval process, in accordance with the requirements of 10CFR50 Appendix B, as committed to in the QATR. C.4 REFERENCES 1. Letter from U.S. Nuclear Regulatory Commission to Niagara Mohawk Power Corporation, dated April 7, 1999,

Subject:

Alternatives for Examination of Reactor Pressure Vessel Shell Welds, Nine Mile Point Nuclear Station, Unit 1 (TAC No. MA4383). 2. Letter from U.S. Nuclear Regulatory Commission to BWRVIP Chairman, dated October 18, 2001,

Subject:

Acceptance for Referencing of EPRI Proprietary Report TR-113596, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74-A)," and Appendix A, "Demonstration of Compliance with the Technical Information Requirements of the License Renewal Rule (10CFR54.21)." 3. NUREG/CR-6260, INEL-95/0045, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," February 1995. 4. Letter from U.S. Nuclear Regulatory Commission to Niagara Mohawk Power Corporation, dated August 11, 1994,

Subject:

Approval of Reduction Factors for Condensation Oscillation Loads in Nine Mile Point Nuclear Station Unit No. 1 (NMP1) Torus (TAC No. M85003). 5. Letter from Niagara Mohawk Power Corporation (NMP1L 1467) to U.S. Nuclear Regulatory Commission, dated September 14, 1999,

Subject:

Submittal of 1999 Inservice Inspection Summary Report and Flaw Indication Evaluations. 6. Letter from Nine Mile Point Nuclear Station (NMP1L 1776) to U.S. Nuclear Regulatory Commission, dated September 19, 2003,

Subject:

Nine Mile Point Unit 1, Docket No. 50-220, Facility Operating License No. DPR Reactor Pressure Vessel Flaw Evaluation. NMP Unit 1 UFSAR Appendix C C-44 Rev. 25, October 2017 7. Letter from U.S. Nuclear Regulatory Commission to BWRVIP Chairman, dated July 28, 1998,

Subject:

Final Safety Evaluation of the BWR Vessel and Internal Project BWRVIP-05 Report (TAC No. M93925). 8. Letter from U.S. Nuclear Regulatory Commission to BWRVIP Chairman, dated March 7, 2000,

Subject:

Supplement to Final Safety Evaluation of the BWR Vessel and Internal Project BWRVIP-05 Report (TAC No. MA3395). NMP Unit 1 UFSAR Appendix C C-45 Rev. 25, October 2017 TABLE C-1 COMMITMENTS Item Commitment Source Schedule 1 Incorporate Appendix A1 into the UFSAR.

• LRA Section A.0 Completed 2 In accordance with 10CFR54.21(b), during NRC review of this application, provide an annual update to the application to reflect any change to the current licensing basis that materially affects the contents of the LRA.
• LRA Section 1.2.1 Completed - Letters dated December 20, 2005, and March 23, 2006 3 Apply for relief from reactor vessel circumferential weld inspections for the period of extended operation. Supporting analyses, procedural controls, and operator training will be completed prior to the period of extended operation to support and confirm that the RPV circumferential weld failure probability remains acceptable for the period of extended operation.
• LRA Section 4.2.3
• LRA Appendix A.1.2.1.3 Completed 4 Supporting analyses will be completed prior to the period of extended operation to confirm that the failure probabilities for the limiting RPV axial welds remain bounded for the period of extended operation.
• LRA Section 4.2.4
• LRA Appendix A.1.2.1.4 Completed 5 For those locations where additional fatigue analysis is required to take advantage of the implicit margin, and to more accurately determine CUF, the EPRI FatiguePro fatigue monitoring software will be implemented prior to the period of extended operation.
• LRA Section 4.3
• LRA Appendix A.1.2.2
• LRA Appendix B.3.2 Completed 6 For the critical reactor vessel components locations shown in Table 4.3-3 of the LRA, additional usage will be added to the baseline CUF using one of the methods described in Section 4.3 of the LRA.
• LRA Section 4.3.1
• LRA Appendix A.1.2.2.1 Completed 7 Transients contributing to fatigue usage of the feedwater nozzles will be tracked by the FMP, with additional usage added to the baseline CUF using the stress-based fatigue method described in Section 4.3 of the LRA.
• LRA Section 4.3.3
• LRA Appendix A.1.2.2.2 Completed 8 Develop a baseline CUF for the specified portions of the following systems: 1.Feedwater/HPCI; 2. Core spray; 3. RWCU (piping inside the RCPB); and 4. Reactor recirculation (and associated shutdown cooling systems lines). If the baseline CUF for a specified portion of a system exceeds 0.4, the limiting locations may require additional monitoring to demonstrate compliance over the period of extended operation.
• LRA Section 4.3.4
• LRA Appendix A.1.2.2.3 Completed NMP Unit 1 UFSAR Appendix C C-46 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 9 Assess the impact of the reactor coolant environment on a sample of critical component locations, including locations equivalent to those identified in NUREG/CR-6260, as part of the FMP. These locations will be evaluated by applying environmental correction factors (Fen) to existing and future fatigue analyses.
• LRA Section 4.3.6
• LRA Appendix A.1.2.2.5
• LRA Appendix B.3.2 Completed 10 The FMP will track transients specific to the emergency cooling system with additional usage added to the baseline CUF for the emergency condensers as described in Section 4.3 of the LRA.
• LRA Section 4.3.7
• LRA Appendix A.1.2.2.6 Completed 11 Enhance the FMP to: 1. Ensure that fatigue usage of the torus-attached piping and other torus locations does not exceed the design limits, add ERV lifts as a transient to be counted by the FMP; and 2. Add the two highest usage torus-attached piping locations, the 12-in core spray suction line for core spray pump 111 that enters the torus at penetration XS-337, and the 3-in containment spray line that enters the torus at penetration XS-326 as fatigue monitoring locations.
• LRA Section 4.6.2
• LRA Appendix A.1.2.4.2
• LRA Appendix B.3.2 Completed 12 The RPV weld flaw evaluations will be revised to consider additional fatigue crack growth and the effects of additional irradiation embrittlement (for beltline materials) associated with operation for an additional 20 yr (i.e., out to at least 46 EFPY) and submitted for NRC review and approval no later than 2 yr prior to the period of extended operation. If the revised calculation shows the identified flaws cannot meet the applicable acceptance criteria, the indications will be reexamined in accordance with ASME Section XI requirements.
• LRA Section 4.7.4
• LRA Appendix A.1.2.5.1 Completed 13 Enhance the BWRVIP to address: 1. BWRVIP-18 open item regarding the inspection of inaccessible welds for core spray system. As such, Nine Mile Point will implement the resolution of this open item as documented in the BWRVIP response and reviewed and accepted by the NRC; 2. The inspection and evaluation guidelines for steam dryers are currently under development by the BWRVIP committee. Once these guidelines are documented and reviewed and accepted by the NRC, the actions will be implemented in accordance with the BWRVIP program; 3. The baseline inspections recommended in BWRVIP-47 for the BWR lower plenum components will be incorporated into the appropriate program and implementing documents; and 4. The reinspection scope and frequency for the grid beam going forward will be based on BWRVIP-26A guidance for plant-specific flaw
• LRA Appendix B.2.1.8 Completed NMP Unit 1 UFSAR Appendix C C-47 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule analysis and crack growth assessment. The maximum reinspection interval for the grid beam will not exceed 10 yr consistent with standard BWRVIP guidance for the core shroud. The reinspection scope will be equivalent to the UT baseline 2005 inspection scope. In addition, the reinspection scope will include an EVT-1 sample inspection of at least two locations with accessible indications within the initial 6 yr of the 10-yr interval. The intent of the EVT-1 is to monitor the known cracking to confirm flaw analysis crack growth assumptions. 14 Enhance the OCCWS Program to: 1. Ensure that the applicable commitments made for GL 89-13, and the requirements in NUREG-1801, Section XI.M20, are captured in the implementing documents for GL 89-13, "Service Water System Problems Affecting Safety Related Equipment Program Plan;" 2. Incorporate into the OCCWS Program the requirements of NUREG-1801, Section XI.M20, that are more conservative than the GL 89-13 commitments; and 3. Revise the preventive maintenance and heat transfer performance test procedures to incorporate specific inspection criteria, corrective actions, and frequencies.
• LRA Appendix B.2.1.10 Completed 15 Enhance the CCCWS Program to: 1. Expand periodic chemistry checks of the system consistent with the guidelines of EPRI TR-107396; 2. Direct periodic inspections to monitor for loss of material in the piping of the CCCWS; 3. Implement a Corrosion Monitoring Program for larger bore CCCW piping not subject to inspection under another program; 4. Establish the frequencies to inspect for degradation of components in CCCWS, including heat exchanger tube wall thinning; 5. Perform a heat removal capability test for the control room HVAC system at least every 5 yr; 6. Establish periodic monitoring, trending, and evaluation of performance parameters for the RBCLC and control room HVAC; 7. Provide the controls and sampling necessary to maintain water chemistry parameters in CCCWS within the guidelines of EPRI Report TR-107396; and 8. Ensure acceptance criteria are specified in the implementing procedures for the applicable indications of degradation.
• LRA Appendix B.2.1.11 Prior to period of extended operation NMP Unit 1 UFSAR Appendix C C-48 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 16 The Boraflex Monitoring Program will be enhanced to: 1. Require periodic neutron attenuation testing and measurement of boron areal density to confirm the correlation of the conditions of test coupons to those of Boraflex racks that remain in use during the period of extended operation; and 2. Establish monitoring and trending instructions for in-situ test results, silica levels, and coupon results.
• LRA Appendix B.2.1.12 Completed 17 Revise applicable procedures related to the Crane Inspection Program to add specific direction for performance of corrosion inspections, with acceptance criteria, for certain hoist-lifting assembly components.
• LRA Appendix B.2.1.13 Completed 18 Enhance the Compressed Monitoring Program to: 1. Develop new activities to manage the loss of material, stress corrosion cracking, and perform periodic system leak checks; 2. Expand the scope, periodicity, and inspection techniques to ensure that the aging of certain subcomponents of the dryers and compressors (e.g., valves, heat exchangers) are managed; 3. Develop and implement activities to address the failure mechanism of stress corrosion cracking in unannealed red brass piping; 4. Establish activities that manage the aging of the internal surfaces of carbon steel piping and that require system leak checks to detect deterioration of the pressure boundaries; and 5. Expand the acceptance criteria to ensure that the aging of certain subcomponents of the dryers and compressors (e.g., valves, heat exchangers) are managed.
• LRA Appendix B.2.1.14 Completed 19 Enhance the Fire Protection Program to: 1. Incorporate periodic visual inspections of piping and fittings located in a non-water environment, such as Halon and CO2 fire suppression systems components, to detect evidence of corrosion and any system mechanical damage that could affect its intended function; 2. Expand the scope of periodic functional tests of the diesel-driven fire pump to include inspection of engine exhaust system components to verify that loss of material is managed; 3. Perform an engineering evaluation to determine the plant-specific inspection periodicity of fire doors.
• LRA Appendix B.2.1.16 Completed NMP Unit 1 UFSAR Appendix C C-49 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 20 Enhance the Fire Water System Program by revising applicable existing procedures to: 1. Incorporate inspections to detect and manage loss of material due to corrosion into existing periodic test procedures; 2. Specify periodic component inspections to verify that loss of material is being managed; 3. Add procedural guidance for performing visual inspections to monitor internal corrosion and detect biofouling; 4. Add requirements to periodically check the water-based fire protection systems for microbiological contamination; 5. Measure fire protection system piping wall thickness using non-intrusive techniques (e.g., volumetric testing) to detect loss of material due to corrosion; 6. Establish an appropriate means of recording, evaluating, reviewing, and trending the results of visual inspections and volumetric testing; 7. Define acceptance criteria for visual inspections and volumetric testing; and 8. Develop new procedures and PM tasks to implement sprinkler head replacement and/or inspections to meet National Fire Protection Association (NFPA) 25, "Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems," Section 5.3.1 (2003 edition) requirements.
• LRA Appendix B.2.1.17 Completed 21 Enhance the Fuel Oil Chemistry Program to: 1. Establish a requirement to perform quarterly trending of water and sediment; 2. Provide guidelines for the appropriate use of biocides, corrosion inhibitors, and/or fuel stabilizers to maintain fuel oil quality;
• LRA Appendix B.2.1.18 Prior to period of extended operation 3. Add requirements to periodically inspect the interior surfaces of the emergency diesel fuel oil storage tanks for evidence of significant degradation, including a specific requirement that the tank bottom thickness be determined by UT or other industry-recognized methods; 4. Add a requirement for quarterly trending of particulate contamination analysis results; 5. Ensure acceptance criteria are specified in the implementing procedures for the applicable indications of potential degradation; 6. Establish a requirement for periodic opening of the diesel fire pump fuel oil day tank drain; and 7. Establish a requirement to remove water, if found.
• LRA Appendix B.2.1.18 Completed NMP Unit 1 UFSAR Appendix C C-50 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 22 Enhance the Reactor Vessel Surveillance Program to: 1. Incorporate the requirements and elements of the ISP, as documented in BWRVIP-116 and approved by NRC, or an NRC-approved plant-specific program, into the Reactor Vessel Surveillance Program, and include a requirement that if Nine Mile Point surveillance capsules are tested, the tested specimens will be stored in lieu of optional disposal. When the NRC issues a final SER for BWRVIP-116, Nine Mile Point will address any open items and complete the SER action items. Should BWRVIP-116 not be approved by the NRC, a plant-specific Reactor Vessel Surveillance Program will be submitted to the NRC 2 yr prior to commencement of the period of extended operation; and 2. Project analyses of USE and P-T limits to 60 yr using methods prescribed by RG 1.99, Revision 2, and include the applicable bounds of the data, such as operating temperature and neutron fluence.
• LRA Appendix B.2.1.19 Completed 23 Develop and implement a One-Time Inspection Program, which also includes the attributes for a Selective Leaching of Materials Program.
• LRA Appendix B.2.1.20
• LRA Appendix B.2.1.21 Completed 24 Develop and implement a Buried Piping and Tank Inspection Program which includes a requirement that before entry into the period of extended operation, if an opportunistic inspection has not occurred, Nine Mile Point will excavate Unit 1 degradation susceptible areas to perform focused inspections. Upon entering the period of extended operation, Nine Mile Point will perform a focused inspection within 10 yr, unless an opportunistic inspection occurred within this 10-yr period.
• LRA Appendix B.2.1.22 Completed 25 An augmented VT-1 visual examination of the containment penetration bellows will be performed using enhanced techniques qualified for detecting SCC, per NUREG-1611, Table 2, Item 12.
• LRA Appendix B.2.1.23 Completed 26 Enhance the Structures Monitoring Program to: 1. Expand the program to include the following activities or components in the scope of license renewal but not within the current scope of 10CFR50.65: a. The steel electrical transmission towers required for the SBO and recovery paths. 2. Expand the parameters monitored during structural inspections to include those relevant to aging effects identified for structural bolting; and 3. Implement regularly scheduled groundwater monitoring to ensure that a benign environment is maintained.
• LRA Appendix B.2.1.28 Completed NMP Unit 1 UFSAR Appendix C C-51 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 27 Develop and implement a Non-EQ Electrical Cables and Connection Program.
• LRA Appendix B.2.1.29 Completed 28 Enhance the Non-EQ Electrical Cable and Connections Used in Instrumentation Circuit Program to: 1. Implement reviews of calibration or surveillance data for indications of aging degradation affecting instrument circuit performance. The first reviews will be completed prior to the period of extended operation and every 10 yr thereafter; and 2. In cases where a calibration or surveillance program does not include the cabling system in the testing circuit, or as an alternative to the review of calibration results described above, provide requirements and procedures to perform cable testing to detect deterioration of the insulation system, such as insulation resistance tests or other testing judged to be effective in determining cable insulation condition. The first test will be completed prior to the period of extended operation. The test frequency of these cables shall be determined based on engineering evaluation, but the test frequency shall be at least once every 10 yr.
• LRA Appendix B.2.1.30 Completed 29 Enhance the Preventive Maintenance Program to: 1. Expand the PM Program to encompass activities for certain additional components identified as requiring aging management. Explicitly define the aging management attributes, including the systems and the component types/commodities included in the program; 2. Specifically list those activities credited for aging management; 3. Specifically list parameters monitored; 4. Specifically list the aging effects detected; 5. Establish a requirement that inspection data be monitored and trended; and 6. Establish detailed parameter-specific acceptance criteria.
• LRA Appendix B.2.1.32 Completed 30 Enhance the System Walkdown Program to: 1. Train all personnel performing inspections in the Systems Walkdown Program to ensure that age-related degradation is properly identified and incorporate this training into the site Training Program; and 2. Specify acceptance criteria for visual inspections to ensure aging-related degradation is properly identified and corrected.
• LRA Appendix B.2.1.33 Completed NMP Unit 1 UFSAR Appendix C C-52 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 31 Enhance the Non-Segregated Bus Inspection Program to: 1. Expand visual inspections of the bus ducts, their supports and insulation systems; 2. Create new provisions to perform as an alternative to either thermography or periodic low-range resistance checks of a statistical sample of the bus ducts accessible bolted connections, a visual inspection for the connections that are covered with heat shrink tape, sleeving, insulating boots, etc., and 3. Define acceptance criteria for inspection of the bus ducts, their support and insulation systems, and the low-range ohmic checks of connections.
• LRA Appendix B.2.1.34 Completed 32 Develop and implement a Fuse Holder Inspection Program.
• LRA Appendix B.2.1.35 Completed 33 Enhance the Bolting Integrity Program to: 1. The Structures Monitoring, PM, and Systems Walkdown Programs will be enhanced to include requirements to inspect bolting for indication of loss of preload, cracking, and loss of material, as applicable; 2. Include in administrative and implementing program documents references to the Bolting Integrity Program and industry guidance; and 3. Establish an augmented inspection program for high-strength (actual yield strength 150 ksi) bolts. This augmented program will prescribe the examination requirements of Tables IWB-2500-1 and IWC-2500-1 of ASME Section XI for high-strength bolts in the Class 1 and Class 2 component supports, respectively.
• LRA Appendix B.2.1.36 Completed 34 Enhance the Protective Coating Monitoring and Maintenance Program to: 1. Specify the visual examination of coated surfaces for any visible defects includes blistering, cracking, flaking, peeling, and physical or mechanical damage; 2. Perform periodic inspection of coatings in the Drywell every refueling outage versus every 24 months. Inspection of coatings in the Torus will be performed every other refueling outage; 3. Set minimum qualifications for inspection personnel, the inspection coordinator, and the inspection results evaluator; 4. Perform thorough visual inspections in areas noted as deficient concurrently with the general visual inspection; 5. Specify the types of instruments and equipment that may be used for the inspection; 6. Pre-inspection reviews of the previous two monitoring reports before performing the condition assessment; 7. Establishment of guidelines for prioritization of repair areas and monitoring these areas until they are repaired; and
• LRA Appendix B.2.1.38 Completed NMP Unit 1 UFSAR Appendix C C-53 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 8. Require that the inspection results evaluator determine which areas are unacceptable and initiate corrective action. 35 Develop and implement a Non-EQ Electrical Cable Metallic Connections Inspection Program.
• LRA Appendix B.2.1.39 Completed 36 As acknowledged by the NRC, the ASME Code Committee is evaluating the acceptability of roll/expansion techniques as a permanent repair for CRD stub tubes via Code Case N-730. Nine Mile Point will continue to follow the status of the proposed ASME Code case and will implement the final Code case, as conditioned by the NRC, once it has been approved. If the Code case is approved by ASME but not yet listed in RG 1.147, Unit 1 will seek NRC approval of the Code case on a plant-specific basis as conditioned by the NRC. It will be programmatically required that during the period of extended operation, should a CRD stub tube rolled in accordance with the provisions of the Code case resume leaking, Nine Mile Point will implement one of the following zero leakage permanent repair strategies prior to startup from the outage in which the leakage was detected: 1. A welded repair consistent with BWRVIP-58-A, "BWRVIP Internal Access Weld Repair" and Code Case N-606-1, as endorsed by the NRC in RG 1.147. 2. A variation of the welded repair geometry specified in BWRVIP-58-A subject to the approval of the NRC using Code Case N-606-1. 3. A future developed mechanical/welded repair method subject to the approval of the NRC.
• LRA Appendix B.2.1.8 Completed 37 Enhance the program to evaluate component susceptibility to loss of fracture toughness. Assessments and inspections will be performed, as necessary, to ensure that intended functions are not impacted by the aging effect.
• LRA Appendix B.2.1.8 Completed 38 An EVT-1 examination of the Unit 1 feedwater sparger end bracket welds will be added to the BWRVIP. The inspection extent and frequency of the end bracket weld inspection will be the same as the ASME Section XI inspection of the feedwater sparger bracket vessel attachment welds.
• NMP Letter NMP1L 2005, December 1, 2005 Completed 39 The Masonry Wall Program (as managed by the Structures Monitoring Program) will be enhanced to provide guidance for inspecting Unit 1 non-reinforced masonry walls that do not have bracing and are within scope of license renewal more frequently than the reinforced masonry walls.
• NMP Letter NMP1L 2005, December 1, 2005 Completed NMP Unit 1 UFSAR Appendix C C-54 Rev. 25, October 2017 TABLE C-1 Item Commitment Source Schedule 40 Unit 1 will perform an EVT-1 inspection of the thermal shield to flow shield weld starting in 2007 and proceeding at a 10-yr frequency thereafter consistent with the ISI inspection interval.
• NMP Letter NMP1L 2005, December 1, 2005 Completed 41 The NRC review of BWRVIP-76 is not yet complete. When the NRC review of BWRVIP-76 is complete, Nine Mile Point will evaluate the NRC SER and complete the SER action item(s), as appropriate.
• LRA Appendix B.2.1.8 Completed 42 Nine Mile Point will perform volumetric examinations on the Unit 1 drywell shell during the 2007 refueling outage, and an engineering evaluation will be performed to determine the actions necessary for Unit 1 operation through the period of extended operation in accordance with the Drywell Supplemental Inspection Program.
• NMP Letter NMP1L 2037, April 4, 2006 Completed}}